ML20105B648

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Rev 0 to Supplemental Reload Licensing Rept for River Bend Station,Reload 4,Cycle 5
ML20105B648
Person / Time
Site: River Bend Entergy icon.png
Issue date: 08/31/1992
From: Albertson D, Klapproth J
GENERAL ELECTRIC CO.
To:
Shared Package
ML20105B634 List:
References
23A7181, 23A7181-R, 23A7181-R00, NUDOCS 9209210079
Download: ML20105B648 (17)
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1 GENuclear Energy l

US C.ee.' J.ew 23A7181 S r .'ese CA % t:5 g,,;,go, o Class!

- Augu3t 1992 i l

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23A7181, Rev. 0  !

Supplemental Reload Licensing Report

. for i

River Bend Station 4

Reload 4 Cycle 5 a

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l Approved:

  • J. F. Klappr k

anager Approved:

3. .Albe 1 son, Maanger

/m L Fuel Lkensing kaloed Nuclear Engineering t

9209210079 920911-PDR ADOCK 05G00458 P PDR

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Rivor Bend

. Re!M 4 twm Rev o Ist portant Notice Regarding Contents of This Report Please Read Carefully This report was prepared by General Electric Company (GE) solely for Gulf States Utilities Company (GSU) for GSU's use 'vith the U. S. Nucleat Regulatory Commission (USNRC) for amending OSU's operating license of the River Bend Station. The information contained in this report is believed by GE to be an accurata and true representation of the facts known, obtained or provided to GE at the time this report was prepared.

The only undertakings of GE respecting information in this document are contained in the contract between GSU and GE for fuel bundle fabrication and related services 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, by anyone other than GSU for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, neither GE nor any of the contributors to this' document makes any representation or warranty (expressed 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 cwned 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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l River Bend

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l Acknowledgment  !

The engineering and reload licensing analyses which form the technical bs. sis of this Supple-mental Reload Licensing Submittal, were performed by P. K. Wu of Fuel Engineering. The

] Supplemental Reload Licensing Submittal was prepared by P. A. Lambert and verified by J. L Rash of Regulatory and Analysis Services.

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River Band ,

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Refond 4 __ pu o The basis for this report is General Electric Standard Application for Reactor fuel, i

NEDE 240ll P A 10. February 1991; and the U. S. Supplement, NEDE 240ll P A 10 US. March 1991.

j  !. Plant unique items

Appendix A
Analysis Conditions Appendix B: Basis for Analysis of Loss of feedwater Heater Event
2. Reload Fuel Bundles Euel Tvne Cycle Imadid Number 4

Irradiated:

W BP8 SRB 299 (BP8x8R) 2 24 l GE88 PSSOB322 90Z 120M 4WR 150 T(BS322C)(GE8x8EB) 3 95 i

GE8B PSSQB322 8GZ 120M 4WR 150 T(BS:228)(GE8x8EB) 3 128 GE8B P8 SOB 331-llGZ 120M 4WR 150 T(CE8x8EB) 4 16 GESB.PSSQB33310GZ 120M 4WR 150 T(GE8x8EB) 4 160 i

New:

i GE88 P8SQB33410GZ 120M JWR 150 T(GE8x8EB) 5 2000

Total 624 i

3- Reference Core Loading Pattern mwd /ST mwd /NR Nominal previous cycle core average exposure at end of cycle: 20.~75 22,680 Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 20,575 22,680 Assumed reload cycle core average exposure at beginning of cycle: 11,520 12,699 Assumed reload cycle core average exposure at end of cycle: 21,625 23,837 Core loading pattern: Figure 1

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4. Cgculated Core Effective Multiplication and Control System Worth No Volds, 20 C

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Beginning of Cycle, K i Uncontrolled 1.128 d

l Fully controlled 0350

{ Strongest control rod out 0.987

R. Maximum increase in cold core reactivity with exposure j into cycle, AK 0.002 .

1 j 5. Standby Liquid Control System Shutdown Capability _

Boron Shutdown Margin (AK)

! (Ram) (20'C. Xenon Free) l l 660 0.029

6. Reload Unique GETAB AOO Analysis initial Condition Farameters Exposure range: BOC5 to EOC5 i Fuel Pealdna Factors Bundle Power Bnedle Flow initial Design Losal Radial 631al R Factor (MWt)

I (1.000 lb/hr) MCPR t.

GESx3EB 1.20 1.56 1.40 1.051 7.059 108.1 1.13 1

j BP8x8R 1.20 1.55 1.40 1.051 7,049- 108.4 1.13 l

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River Bend nu,n Reload 4 ,

_ nn o 7, Seinted Margins improvement Options 4

Recirculation pamp trip: Yes Rod withdrawallimiter: Yes Thermal power monitor: Yes improved scram time: No

Exposure dependent limits
No Expoure poin;s analyzed: 1 (EOC)
8. Operating Flexibility Options
Single loop operation: Ws

! 1. cad line limit: No

. Extended load line limit: No

, Maximum extended load line limit No Increased core flow throughout the cycle: No Increased core flu at end of cycle: No Flow point analyzed: N/A Feedwater temperature reduction throughout the cycle: No Final feedwater temperature reduction: No- ~

Tempera;ure reduction: N/A ARTS Program: No Maximum extended operating domain: No ADS valve out of service: No

Safety / relief valve out of service
No Main steam : solation valve out of service: No Turbine bypass out of service: No ,

4 EOC Recirculation par.:p trip out of service: No Feedwater heaters out of service: Yes 4

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[ 9. Cocu wide AOO Analysis Results "

i -- Methods used: GEMINI and GEXL PLUS -

_UnntrK!fd.1CffL i

Flux Q/A

... Event & NRR) (% NBR)- GE8 DEB BP8 8R Ugutt h

i Exposure range: BOCS to EOC5 i

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Loa-J rejection without bypass 298. 103 0.06 0.06 -2 i.

Feedwater controller faliore 219 108 0.05 0.05 3 1

j. Pressure regulator fadure 144 104 0.03 0.03 T i downscale i

F Loss of FO'F feedwater *

  • 0 11- *

! 0.11

heating i

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10. local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary j'  :
- The generic bounding BWR/6 rod withdrawal error analysis described in NEDE.24011.P.A 10.US is applied; the resulting ACPR is 0.11. t t

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11. Cycle MCPR Values
  • l-Safety limit: 1.07

$ Single loop operation safety limin 1.08 l

z Non.oressurir.ation events:

j GE8XRER BPRXRR i

_ Exposure range
- BOCS to -EOC5 4

{ Loss of 100* F feedwater heating 1.18 1.18-i i Rod withdrawal error 1.18 - 1.18 l

j Fuelloading error" 1.22 --.

l Pressurhatta.nntnis:

1 GEEX8E8 BPRX8R-i Exposure range: BOC5 to EOC5' i- Load rejection without bypass -

1.0 - - 1.1.F 4

Feedwater controller tailure . 1.12 1.12' i

j Pressure regu'ator failure downscale - 1.10- 1.10 1

i p 12. Overpressurization Analysis Summary 4 .

Put 'Pr Event _. (psig) (paig) _ Plant Resoonse
MSIV closure : 1212 1256 Figure 5 '

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{ (flux scram) '

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  • 0EMINI ODYN adjustment fae: ors are provided in the letter from J. S. Charnley (GE) to _

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2 M. W. Hodges (N RC), GEMINI ODYN Adjustment Factorsfor BWR/6, dated July 6,1937. The -

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'  : limiting transients for River Bend Station, Cycle 5, are rod withdrawal error and loss of 100* F -

feedwater heating. .

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"See lerter, J. F. Klapproth (G E) te R. C, Jones, Jr. (NRC). Rotated B4mdle Evaluation, July 20. - ,

-1992.

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4-4 13 . Loading Ertvr Results*

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Variable water gap misoriented bundle analysis: Yes Even!._ LCfB i.

Misorier.ted fuel bundle 0.15 9

i i 14. Control Rod Drop Analysis Results j

River Bend Station is a banked position withdrawal sequence plant,'therefore, the control rod drop accident analysis is not required. ' NRC approval is documented in l NEDE 240ll P A 10 US, March 1991.

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15. Stability Analysis Results s I

5 I-1 GE SIL-330 recommendations have been included in the River Bend Station operatin j

procedures and Technical Specifications) therefore, the stability ana!ysis is not required NRC approval for deletion of a cycle specific stability analysis is documented in Amendme 8 to NEDE 240ll P A US. . River Bend Statbn recognizes the issuance of NRC Bulletin No l ~

88-07. Supplement 1. Power Osciliations in Boiling Water Reactor. {BWRt),-and i

{

with the recommendations contained therein.

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leaer, J. F. Kist prcth (GE) to R. C. Jones, Jr. (NRC), Rotuted Bundle Eva P.-

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16. Losser<oolant Accident Results*

f LOCA method used: SAFE /REFLOOD (see River Bend Station Final Safe.ty

[ Analys Report)"

.i Bundle Type: GE8B P8SQB33410GZ-120M 4WR 150-T (GE8x8Eli)

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3 Average Planer E3gtsure . _ Mt\PLHGR (kw/ftf a  ;

l l (GWd/ST) (GWd/MT) Most Limithg Lgast Lim

  • ting 4 -

0.0 0.0 11.36 11.90 j ,

0.2 0.2 11.42 11.93 I; 1.0 L1 11.54 -l' 03

,) 2.0 2.2 -11.71- 12.18

i 3.0 3.3 ~ 11.90 i234 l 4.0 4.4 12.09 12.51 5.0 5.5 12.30 12.68 6.0 6.6 12.52 12.86 7.0 7.7 12.74 13.05-i 8.0 8.8 12.97 13.23 9.0 9.9 13.21 13A1

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. 10.0 11.0 13.41 13.56

) 12.5 13.8 13.60 13.65.-

4 15.0 16.5 1131 13.32 20.0 22.0 12.64 12.64 25.0 27.6 - 11.% - 11.97 .

l -35.0 - 38.6 10.46 10.53 ,

45.0 49.6 9.08 9.21
50.0' 55.1- 6.97- 7.04' L

The peak clad temperature (PCT) is s2t 31 *F at all exposures; the local oxidation (fraction) is 50.073 at all exposures. The MAPLHOR multiplier for single loop operatiou (SLO)is 0.83 if---

Technical Sr.4cification D/O start time is greater than 13 seconds and less than or equal to 30

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seconds; 0.84 if Technical Specification D/G start time is 13 seconds or less.

< 'For format explanation, see letter, J. S. Charnley (GE) to' M. W. Hedges (NRC),-

Recommended MAPLHGR Technical Specificationsfor Multiple Lattice Fuel Designs, March 9, .

1987. Most Limiting and Least Limiting refer to the lowest and highest. limits, respectively, of any enriched lattice in the bundle.- .

, "As modilled by GE Report EAS-66-1988, ECCS Evaluation for 30 Second Diesel Generator Starr Tunesfor River Bend Station, (R. U. Fortney and J. L Jacobs), August 1989.

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!:: E M M M M M M M M M M M M M E i!MMMMMMMMMMMMBH L ':!MMMMMMMMMMMME -

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1-3 5 7 91113151719212325272931333537394143454749515355 FUEL TYPE A GE88-PBSQB322-8GZ-120M 4WR-150-T D=GE88 P8508331 11GZ-120M-4WR-150-T B GE88 P85QB322 9GZ-120M-4WR-150-T 'E=BP8 SRB 299 3.C GE0B-P85QB333-10GZ-120M 4 Wit-150-T F=GEBB-P85QB334-10GZ-120M-)WR-150 g Figure 1. Reference Core Loading Pattern em n

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Figure 2. Plant Response to load Rejection without Bypass (BOC5 to EOC5)

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j Appendix A '

Analyses Conditions -

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To reDect actual plant parameters accurately, 'the values shown in Table A 1 were used this

{ cycle.

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. Table A 1 t

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j- Parameter Analysis Vglys l Thermal power,'MWt 2894 1-j Dome presure, psig -1025-4 j- Steam How, M]b/hr 12.45 l Turbine pressure, psig 986-

-Core Dow, Mlb/hr .- 84.5 Reactor pressure, psia 1055

Inlet enthalpy, BTU /lb 527,9. .

4 l Non fuel power. fraction- 0.039--

1-L No. of Dual Mode Safety /Reliet Valves ' 16 '-

Relief mode low setpoint, psig _ 1133--

- Safety mode low setpoint, psig 1177-Capacity,-Ib/hr 1 831,000

(Ref. pressure, psig) . (1080):

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Appendix B i c .

i . Basis for Analysis of Loss of.reedwater Heater Event l'

l- The loss-of feedwater heating event was analyzed at 102% rated power using the BWR ,

j Simulator Code (Reference B 1). The use of this code is permitted in GESTAR II(Reference l B 2). Tne transient plots, neutron Dux and heat flux 9:?ues normally reported .in Section 9 are not an output of the BWR Simulator code; therefore, these items are not included in this document.

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! Referenc.ts B1 Steady-State NuclearMethodr, NEDE-30130 P A and NEDO-30130 A. Apr31985.'

I B-2 General Electric Standard Application for Reactor Fuel, NEDE 24011-P A (latest approved j version).

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