ML20006E647

Supplemental Application for Amends to Licenses NPF-72 & NPF-77,correcting Info Submitted in Re Use of Vantage 5 Fuel or Combination of Vantage 5 Fuel & Optimized Fuel Assembly Core at Plant

ML20006E647
Person / Time
Site:	Braidwood
Issue date:	02/16/1990
From:	Hunsader S COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20006E648	List: ML20006E647 ML20006E649
References
NUDOCS 9002260176
Download: ML20006E647 (9)

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Commonwealth Edison l

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February 16, 1990 j

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b Dr. Thomas E. Murley, Director i

Office of Nuclear Reactor Regulation U.S. Nuclear: Regulatory Commission-

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Washington, D.C.

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, Subjects -Braidwood Station Units 1 and-2 Supplement to Application for Amendment

'to Facility Operating Licenses NPF-72 and NPF-77 HR0 Docket Nos. 50-456 and 50-457

References:

La) October 19, 1989' S.C. Hunsader letter to_T.E. Murley b)' January 9, 1990 S.C. Hunsader letter j

to T.E. Murley Dear Dr. Murley In reference (a) pursuant to 10 CFR 50.90, Commonwealth Edison

. proposed to-amend Appendix A, Technical Specifications of Facility Operating Licenses NPF-72 and NPF-77.= The proposed amendment' requests the use of

' VANTAGE 5 fuel or the combination-of VANTAGE 5 fuel-and the present-Optimized Fuel Assembly core at: Braidwood Station Units 1 and!2. Reference (b) provided additional-information to supplement the No Significant; Hazards evaluation.

l The purpose of-this letter is to provide corrections to information previously submitted..

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Attachment A provides a revised page 24 of Attachment 4 to reference

_a)'to correct the inaximum local Zr-H O reaction percent from 2.14% to 3.26%..

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Included in Attachment B is Section 6.0-which is being provided for l

_4 completeness.

Similar information had been previously_provided and is Lincludedfin Attachment 2 to reference (a).

Attachment.O provides a revised Section 7.0'that corrects the identification numbers for the reference doctunento, listed. Also documents l'

that were previously listed, but are not applicable to the information in the report text, have been deleted.

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2-Included in Attachment D is a revised marked-up Technical Specification page 2-8, that changes " Unit 1, Cycle 4" to " Unit 1, Cycle 3",

and " Unit 2, Cycle 3" to " Unit 2, Cycle 2", both of which are the correct power cycles for Braidwood Station.

Commonwealth Edison is notifying the State of Illinois of this supplement to the application for the amendment by transmitting a copy of this letter and its attachments to.the designated State Official.

1 Please direct any questions regarding this submittal to this office.

Very truly yours, 0

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S. C. Hunsader Nuclear Licensing Administrator

/Imw:0487T:5 Attachments A, B, C & D cc: Braidwood Resident Inspector S.P. Sands - NRR Office of Nuclear Reactor Safety - IDNS S. Sun - NRR f

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B/B-UFSAR A +-+ct cJa m e n t A

The maximum calculated local metal-water was 3 26%%ich is well below-th embrittlement limit lof'17% specified in 10CFR50.46.- The total core' wide _

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- metal water reactions is-less than 0.3% for,all breaks, as compared with the-1% criterion of 10CFR50.46 and in all cases the cladding tosperature transient 4

lwas terminated at'a time when the core gecastry.was still amenable'to cooling. As a result, the core temperature will continue-to drop and thes ability to remove decay heat generated fn the fuel for an extended ~ period.of-L time will be provided.

These results provide assurance that operation with VANTAGE 5 fuel and with the RCS hot leg temperature in the range of 600 to 619.3'F can be acecaplished within the requirements of 10CFR50.46 and Appendix K to 10CFR50.46.

Small Break Results This section presents the results of a spectrue of small break sizes analyzed for the Byron /Braidwood Stations. As noted previously,-the calculated peak clad ~ temperature resulting from a small break LOCA is less than that calculated for.a large break.

Based on the results of LOCA sensitivity.

studies (Reference 14 and 21) the limiting small break was found.to be.less than a 10-inch diameter rupture of the RCS cold leg.

The worst breaks size (small break) is a 3-inch diameter break in the cold leg. This limiting break size was also analyzed for the reduced RCS operating temperatures to show that the reduced temperature results in a less severe transient.

The time sequence of events and the results for all the breaks analyzed is shown-in Tables 15.6-1 and 15.6-4.

During the earlier part of the small break transient, the effect of the break flow is not strong enough to overcome the flow maintained by the reactor coolant pumps through the core as they are coasting down following reactor trip.

Therefore, upward flow through the core is maintained.

The resultant k

heat transfer cools the fuel rods and cladding to very near the coolant I

temperature as long as the core remains covered by a two phase mixture.

This effect is evident in the accompanying figures, d

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6.0

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES Table 6.1 presents a list of the Technical Specification changes and justification for the changes.

The changes noted in Table 6.1 are given in the proposed Technical Specification page mark ups (see Appendix A of this.

report).

Included.in Appendix A are separate mark ups for the Byron and Braidwood Stations Units 1 and 2.

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TABLE ~6.1 Summary and Justification for Byron and Braidwood Stations Units 1 and 2 Technical Snecification Chances for VANTAGE 5 Fuel EAGE-SECTION DESCRIPTION OF CHANGE JUSTIFICATION-q 2-8 Table 2.2-1 Revised the F(AI)

These changes are due to-offset wings and gains the VANTAGE 5 fuel l

with cycle specific

design, identification.

B 2-1 2.1.1 Added DNB correlations These changes reflect B-2-2 Basis and design and Safety

.the DNB correlations Analysis DNBR limits for and the values for the VANTAGE 5 fuel.

F$Hfor the Added new F VANTAGE 5 and 0FA fuel.

gg values.

3/4 1-4 3.1.1.3 BOL deleted from MTC LCO These changes reflect 3/4 1-5 and Surveillance increasing HTC with burnup 4.1.1. 3.-

modified to before decreasing toward i

compare BOL MTC with EOL for VANTAGE 5 core and predicted HTC with to allow entry into Modes 1 burnup and develop rod and 2, if the requirements i

withdrawal limits to of the Action Statements keep HTC negative.

Added are met.

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" Provisions of Specifica-tion 3.0.4 are not applicable." to the Action-Statement.

3/4 1-19 3.1.3.4 Revised the rod drop This change is the time.to 1 2.7 seconds result of an increase and added cycle specific in the core hydraulic identification.

resistance due to the VANTAGE 5 fuel design.

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-r EAGE SECTION DESCRIPTION OF CHANGE JUSTIFICATION 3/4J2-4 3.2.2 Added new F limit ~and This change reflects g

cycle specific the value for-F9 identification.

assumed in the safety analysis for the_

VANTAGE 5 fuel design.

3/4 2-5 3.2.2 Replaced Figure 3.2-2 This curve is consistent-with 2 segment curve.

with the VANTAGE 5 analysis.

3/4 2-7 3.2.2 In 4.2.2.2.f.3 The VANTAGE 5 fuel assemblys add "(except VANTAGE 5 IFM grids.wil not signifi-fuel assembly IFM grids)".

cantly distort the indi-cated flux during the Fxy. surveillance.

RevisedtheF$H.

Thesechangegreflectthe 3/4 2-81 3.2.3

limits, values for FAH assumed in the safety analyses for VANTAGE 5 and 0FA' fuel.

B 3/4 1-2 3/4 1.1.3 Reworded Surveillance This change reflects Basis justification paragraph.

increasing HTC with burnup before decreasing toward EOL for VANTAGE 5 core.

B 3/4 2-1 3/4.2 Revised basis discussion These changes reflect the Basis of DNB.

new DNB correlations used for the VANTAGE 5 and 0FA fuel.

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TABLE 6.1 (continued)

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SECTION DESCRIPTION OF CHANGE JUSTIFICATION

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Changeo the axial This change reflects the-B 3/4 2-1 3/4.2.l Basis peaking factor

'value for F assumed in 9

multiplier to F limit.-

the safety. analyses for n

either 0FA or VANTAGE 5 fuel design.

B 3/4'2-4 3/4.2.2 Revised basis discussion This change reflects the 3f 3/4.2.3 for rod bow penaity.

new DNB correlations used-for VANTAGE 5 and 0FA' fuel.

Revigedbasisdiscussion RevisedF$H B 3/4 2-5 3/4.2.2 limits-

~3/4.2.3 of F limits.

to' include VANTAGE 5 AH fuel-design, i,

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7.0 REFERENCES

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1.

Davidson, S.

L.,

loril, J. A., " Reference Core Report - 17x17 Optimized Fuel Assembly," WCAP-9500-A,.May 1982.

2.

Davidson, S. L.. Kramer, W. R. (Eds. ) " Reference Core Report VANTACE 5 Fuel Assembly," WCAP-10444-P-A, September 1985.

'3.

'Davidson, S. L. (Ed.), et al., " Westinghouse Reload Safety Evaluation Methodology," WCAP-9272-P-A, July 1985.

4.

Miller, J.

V., " Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," WCAP-8720 (Proprietary), October 1976.

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5.

Weiner, R.

A., et al., " Improved Fuel Performance Models'for Westinghouse-Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A, August 1988.

6.

Davidson, S. L. (Ed.) et al., " Extended Burnup Evaluation of Westinghouse Fuel," WCAP-10125-P-A, December 1985.

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'7.

Letter from W. J. Johnson (W) to M. W. Hodges (NRC), "VANTACE 5 Bottom

.j Nozzle," NS-NRC-88-3366, September 1988.

8.

Skaritka,. J., (Ed. ), Fuel Rod Bow Evaluation," WCAP-8691 Revision 1 (Proprietary), July 1979.

9.

Davidson, S..L.,

Iorii, J. A. (Eds.) " Verification Testing and Analyses of.

the 17x17 Optimized Fuel Assembly," WCAP-9401-P-A, August 1981.

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'10. Skaritka, J.,

et al.

" Westinghouse Wet Annular Burnable Absorber Evaluation Report," WCAP-10021-P-A, Revision 1, October'1983.

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11. Chelemer, H., Boman, L.

H., Sharp, D.

R., " Improved Thermal Design

)

Procedure," WCAP-8567-P-A, February 1989.

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12. Letter from NRC to Westinghouse from Stolz to Eiche1dinger, SER'on-l WCAP-7956,:8054, 8567 and 8762 dated April 1978.

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13. Motley, F.

E., et al., " Westinghouse Correlation WRB-1 for Predicting i

i-Critical Heat Flux in Rod Bundles with Mixing Vane grids," WCAP-8762-P-A and WCAP-8763-A, July 1984.

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14. Letter from E. P. Rahe (W) to Miller (NRC) dated March 19, 1982, NS-EPR-2573, WCAP-9500 and WCAPS 9401/9402 NRC SER Mixed Core j

Compatibility Items.

15. Letter from C. O. Thomas (NRC) to Rahe (W) - Supplemental Acceptance No. 2 j.

.for Referencing Topical Report WCAP-9500, January 1983.

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+q s-REFERENCES (Continued)

16. Letter from W. J. Johnson (W) to M. W. Hodges (NRC), NS-NRC-87-3268

" VANTAGE 5 DNB Transition Core Effects," October 2, 1987.

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17. Letter from M. W. Hodges (NRC) to W. J. Johnson (H), NRC SER on. VANTAGE 5 j

Transition Core Effects, dated February 24, 1988.

18. Schueren, P., McAtee, K. R.,~" Extension of Methodology for Calculating Transition-Core DNBR Penalities,".WCAP-11837, May 1988.-

19. Butler, J. C. and D. S. Love, "Steamline Break Mass / Energy Releases for Equipment Environmental Qualification Outside Containment," WCAP-10961-P (Proprietary) and WCAP-11184 (Non-Proprietary), October 1985.

20. DiTommaso, S. D. et al.. " Byron /Braidwood THot Reduction Final Licensing' Report," WCAP-11386, Rev. 2 (Proprietary) and WCAP-11387, Rev. 2 (Non-Proprietary), November 1987.

21. Kabadi, J. N. et al., "The 1981 Version of the Westinghouse ECCS, Evaluation Model Using the BASH Code," WCAP-10266 P-A, Revision 2, with Addenda, (Proprietary), March 1987.

22. Besspiata, J. J. et al., "The 1981 Version of the Westinghouse ECCS, Evaluation Model Using the BASH Code, Power Shape Sensitivity Studies,"

WCAP-10266-P-A, Revision 2, Addendum 1, (Proprietary), December 15, 1987.

23. Lee, N., et al., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A.(Proprietary) and WCAP-10081-A (Non-Proprietary), August 1985.

24. Young, M., et al., "BART-1A: A Computer Code for the Best Estimate Analyzed.Reflood Transients," WCAP-9561-P-A, 1984.

25. Meyer, P.E., "NOTRUMP, A Nodal Transient _Small Break and General Network Code," WCAP-10079-P-A (Proprietary) and WCAP-10080-P-A (Non-Proprietary)

August 1985 0676T 7-2