ML20154L373

Forwards Request for Addl Info Re Thot Reduction Program & Proposed Amends to Licenses NPF-37,NPF-66,NPF-72 & NPF-75 Revising Tech Spec Re Heat Flux Hot Channel Factor as Function of Core Height

ML20154L373
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	09/16/1988
From:	Olshan L Office of Nuclear Reactor Regulation
To:	Bliss H COMMONWEALTH EDISON CO.
References
NUDOCS 8809260160
Download: ML20154L373 (8)

Text

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Docket Nos.: 50-454, 50-455, 50-456 and 50-457 Mr. Henry E. Bliss Nuclear Licensing Manager Comonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690 1

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - THOT REDUCTION PROGRAM FOR BYRON /8RAIDWOOD By letter dated December 4,1987, you provided a description of your That reduction program and proposed an amendment to ths Byron /Braidwood Technical Specification revising the figure which depicts the normalized heat flux hot chennel factor as a function of core height. Enclosed is a request for additional information that we need to complete our review of your submittal.

Please respond within 45 days of receipt of this letter.

The reporting and/or rccordkeeping requirements contained in this letter affect fewer than ten respondents; therefore OMB clearance is not required under P.L.96-511.

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Leonard N. Olshan, Project Manager Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects cc: See next page DISTRIBUTION:

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Docket Nos.: 50-454, 50-455, 50-456 and 50-457 Mr. Henry E. Bliss Nuclear Licensing Manager Comonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

Dear Mr. Bliss:

SUBJECl: REQUEST FOR ADDITIONAL IhFORMATION - TH0T REDUCTION FROGRAM FOR BYRON /BRAIDWOOD By letter dated December 4,1987 you provided a description of your Thot reduction program and proposed an amendment to the Byron /Braidwood Technical Specification revising the figure which depicts the normalized heat flux hot channel factor as a function of core height. Enclosed is a request for additional infonnation that we need to complew our review of your submittal.

Please respond within 45 days of receipt of tMs letter.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; tnerefore, OMB :learance is not requira1 under P.L.96-521.

AU b h Leonard N. Olsban Project Manager  :

Project Directorate III-2 Division of Reactor Projects - III,

, IV, Y and Special Projects cc: See next page l

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t Mr. Henry Bliss Connonwealth Edison Company Byron /Braidwood CC:

Mr. William Xertier Dr. Pruce von Zellen Atomic Power Distribution Department of Biological Scient:es

, Westinghouse Electric Corporation Northern Illinois University l

Post Office Box 355 DeKalb, Illinois 61107 Pittsburgh, Pennsylvania 15230 Joseph Gallo. Esq. U. S. Nuclear Regulatory Comission Hopkins and Sutter Byron / Resident Inspectors Office 1050 Connecticut Av6., N.W. 4448 North German Church Road Suite 1250 Byron, Illinois 61010 Washington, D. C. 20036 C. Allen Bock, Esquire Rt. 1, Box 182 Post Office Box 342 Manteno. Illinois 60950 Urbana, Illinois 61801 Mrs. Phillip 2. Johnson Regional Administrator 1907 Stratford Lane U. S. hRC, Region III Re:kford, Illinois 61107 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Illinois 60137 Douglass Cassel, Esq.

109 N. Dearborn Street Ms. Bridget Little Rorem $uite 1300 Appleseed Coordinator Chicago, Illinois 60602 117 North Linden Street l Essex, Illinois 60935 Ms. Pat Morrison 5568 Thunderidge Drive Mr. Edward R. Crass Rockford, Illinois 61107 Nuclear Safeguards and Licensing Division David C. Thomas Esq.

Sargent & Lundy Engineers 77 S. Wacker Drive 55 East Monroe Street Chicago, 1111 noir (0601 Chicago, Illinois 60603 Elena Z. Kezelis, E.<q.

U. S. huelear Regulatory Comission Isham, Lincoln & Beale Resident Inspectors Office Three First National Plaza RR#1, Box 79 Suite 5200 Braceville Illinois 60407 Chicago, Illinois 60602 i

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. j Mr. Henry E. Eliss Commonwealth Edison Company - 2- Byron /Braidwood cc:

Mr. Charles D. Jones Director Illinois Emergency Services and Disaster Agency 110 East Adams Strees Springfield, Illinois 627D6 Mr. Michael C. Parker, Chief Division of Engineering Illinois Department of Nuclear Safety .

1035 Outer Park Drive Springfield, l'i11nois 62704 Michael Miller, Esq.

Sidley and Austin One First National Plaza

• Chicago, Illinois 60603 George L. Edgar Newman & Holtzinger, P.C.

1615 L Street, h.W.

Washington, D.C. 20036 Commonwealth Edison Company Byron Station Manager 4450 North German Church Road Byron, Illinois 61010' G

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-l QUESTIONS ON T REDUCTION FOR

, H0T BYRON /BRAIDWOOD UNITS 1 AND 2

1. In Section 3.6 of Attachcent C of the above Reference, which pertains to the impact of the T reduction program on the Technical Specifications, hot you state that "A statistical setpoint study perfonned previously for Byron Units 1 and 2 provided increased margin in total allowance to l various Technical Specification related instrumentation setpoints. Based on an evaluation of the sensitivities of the study for the reduced temperature parameters, it has been concluded that the setpoint allowances l accounted for in the statistical evaluation remain valia."

l Please provide the references for the stated statistical setpoint study and the evaluation. Was this reviewed and approved by the NRC? What is the value of the increased margin in total allowance to the various l Technical Specification related instrument setpoints? Provide the l summary of the study for the reduced temperature parameters for which you i state that the statistical evaluation remains valid.

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2. You have stated in Section 3.1 of Attachment C of the above Reference that both the small break and large b7ak LOCA conditions have been reanalyzed and tha' the reanalyses are presented in Appendix A of your report. Appendix A contains marked-pr. actions of Chapters 6.2.1.5 and 15.6.4 of the Byron /Braidwood FSAR. For the large break LOCA you state, in insert 4 that the chopped cosine power shape results in the most severe calculated consequence as required for LOCA analysis in 10 CFR Part 50, Appendix K. For the small break LOCA you state on page 15.6-19 that "Figure 15.6-48 presents the hot rod power shape utilized to perfor.n the small break analysis presented here. Tf's power shape was chosen became it provides an appropriate distr don of power versus core height and also local power is maximized in the uoper regions of the reactor core (10 ft. to 12 ft.). This power shape is skewed to the top of the core with the peak local power occurring at the 10.0 ft. core elevation."

2 Please explain how the hot rod power shape was arrived at for both the large and small break LOCA to satisfy the requirements of 10 CFR 50.46, Appendix K which states that ". . . A range of power distribution shapes and peaking factors representing power distributions that may occur over the core lifetime shall be studied and the one selected should be that which results in the most severe calculated consequences . . ."

3. In Attachment C, Appendix B of the above Reference, the Technical Specification change for Figure 3.2.-2 is provided. This figure shows K(Z) normalized F (Z) as a function of core height. In com;ents to this  ;

9 revision you have stated in Attachment D of Reference 1 that "This revision to the third line segment of the K(Z) curve will allow reactor operation with an increased heat flux hot channel factor at high core location."

In order to compare the results of your analysis with the revised K(Z) curve plesse provide figures similar to Figure 3.2-2 with the rasults of l your analysis imposed for the power shape. This should include curves of

) linear heat generation rate (kw/ft) vs. elevation (ft) including core l average and hot rod values and the K(Z) limit for the SBLOCA.

4. On P.6 is a statement that the increases in Fg and F g were addressed only in the LOCA analysis. Will these increases be addressed for other accident analyses where they may affect (1) local power density, (2) j minimum DNBR?

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5. In Section 3.4 (P.13) on Non-LOCA Transients - Please explain why the parameters modified differ as shown below:

19.3'F reduction in nominal RCS T,y, (P.13) 19.6*FinTable2.1-1(P.5) 4

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, 6. In Figure 3.4-1 (P.31) - No labels are shown on the curves which show solid and dashed lines for curves of delta 1 vs. T,y,. Please identify what the solid and dashed lines represent. .

7. In Table 15.6-1 (Sheet 1 of 3) P. 15.6-3, why is the event of "rods begin to drop" (42.6 sec) listed after the event of "minimum DNBR occurs" (43.7 sec).

8. In Table 15.6 (P.15.6-35), "Input parameters used in the ECCS analysis,"

there are listings for initial loop flow, inlet and outlet temperature hot values are given. What are and steam pressure. Reduced and nominal T the nominal and reduced Tg values. Provide a background for the values l stated and explain why the nominal values differ from the crossed out values used initially. Is this due to increases in peak linear power and peaking factors Fq and FZ? Was the original analysis with no steam generator tube plugging? Does the 18'F reduction in TH require such a large steam pressure reduction of about 200'F from 977'F7 Is the safety injection flow input for the ECCS analysis reduced by 5% as mentioned in page 67 ,

9. In Table 15.6-3, Page 15.6-36a - Large Break LOCA Results Fuel Cladding l

. Data:

For the column heading Nominal Tgog why is the temr , ;re 622.3*F instead of 618.4*F7 Is the max ECCS column the ECt,S flow without the 51 l l

reduction in safety injection flow as stated in page 6 for 1.0CA accidents?

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10. In Table 15.6-4, page 15.6-37, - SBLOCA Results Fuel Cladding Data:

Why is the T hot = 622.3*F instead of 618.4*F as used originallyi i

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11. On page 15.6-19, why is there a decrease in the elevation for peak power from 10.5 ft. to 10.0 ft.?

12. Have any other Westinghouse plants had a:

a) Mod of T Hot reduction? Which ones?

b) hcd of 3rd line segment of X(Z) curve removed? Which ones?

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