ML20133E154

Requests Addl Info Re TS Change Request 96-01 on Conversion to Cogema Fuel for Plant.Response Requested within 30 Days of Ltr Receipt

ML20133E154
Person / Time
Site:	Sequoyah
Issue date:	01/08/1997
From:	Hernan R NRC (Affiliation Not Assigned)
To:	Kingsley O TENNESSEE VALLEY AUTHORITY
References
TAC-M95144, TAC-M95145, NUDOCS 9701100221
Download: ML20133E154 (8)

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, January 6,1997 Mr. Oliver D. Kingsley, Jr. I President, TVA Nuclear and Chief Nuclear Officer Tennessee Valley Authority 6A Lookout' Place 1101 Market Street

' Chattanooga, TN 37402-2801

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION - TECHNICAL SPECIFICATION CHANGE REQUEST 96-01 ON CONVERSION TO COGEMA FUEL - SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TAC NOS. M95144 AND M95145)

Dear Mr. Kingsley:

The staff is reviewing the subject proposed Technical Specification change request submitted by the Tennessee Valley Authority (TVA) in.its letter dated April 4, 1996. The basis for a proposed no significant hazards consideration determination was published in the Federal Reaister on May 8, 1996 (61.FR j 20856). The staff also met with TVA and the fuel manufacturer, Framatome i Technologies, in a public meeting in Rockville, Maryland on June 20, 1996, to  !

discuss various technical issues regarding this change request.

During our review process, we have identified the need for additional information in order to complete our review of the TVA request. Our request for additional information (RAI) is attached. The schedule for timely completion of this review requires that TVA provide a response to the RAI within 30 days of receipt of this letter.

Please contact me at (301) 415-2010 if you have any questions.

Sincerely, Originat signed by I

.)

Ronald W. Hernan, Senior Project Manager.

Project Directorate 11-3 J Division of Reactor Projects - I/II I Office of Nuclear Reactor Regulation  !

Docket Nos. 50-327 and 50-328

Enclosure:

Request for Additional Information cc w/ enclosure: See next page M( l Distribution:

-Docket, File; PUBLIC SQN Rdg. File S. Varga  !

J. Zwolinski OGC ACRS E. Merschoff, RII l C. Jackson M. Shannon, RII l 9701100221 970108 ]

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P PDR gh k,g f j DOCUMENT NAME: G:\SQN\95144.RAI j TO GET e sepy of thle deeuenent,indcate les the ben: "C" = Copy without attachment / enclosure

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g 4 UNITED STATES E

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20065 0001

. . . . . *# January 8, 1997 Mr. Oliver D. Kingsley, Jr.  !

President, TVA Nuclear and Chief Nuclear Officer  !

l Tennessee Valley Authority l 6A Lookout Place l 1101 Market Street Chattanooga, TN 37402-2801 l

SUBJECT:

l REQUEST FOR ADDITIONAL INFORMATION - TECHNICAL SPECIFICATION

! CHANGE REQUEST 96-01 ON CONVERSION TO COGEMA FUEL - SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 (TAC NOS. M95144 AND M95145)

Dear Mr. Kingsley:

The staff is reviewing the subject proposed Technical Specification change request submitted by the Tennessee Valley Authority (TVA) in its letter dated 1

April 4, 1996. The basis for a proposed no significant hazards consideration determination was published in the Federal Reaister on May 8, 1996 (61 FR 20856). The staff also met with TVA and the fuel manufacturer, Framatome Technologies, in a public meeting in Rockville, Maryland on June 20, 1996, to l discuss various technical issues regarding this change request.

During our review process, we have identified the need for additional information in order to complete our review of the TVA request. Our request for additional information (RAI) is attached. The schedule for timely completion of this review requires that TVA provide a response to the RAI within 30 days of receipt of this letter.

Please contact me at (301) 415-2010 if you have any questions.

Sincerely, b

h, ,M1 -

Ronald W. Hernan, Senior Projecc Manager Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosure:

Request for Additional Information cc w/ enclosure: See next page

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REQUEST FOR ADDITIONAL INFORMATION )

TENNESSEE VALLEY AUTHORITY SE000YAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NUMBERS 50-327. 50-328 l 1. Revision 2 to the BAW-10168 evaluation model was modified during the l staff review with regard to the Moody break flow model and discharge l coefficient. Please verify that the analysis that was used to support j this Sequoyah Nuclear Plant (SQN) fuel change and Technical i

Specification (TS) amendment was performed using the approved model.

2. The proposed TSs deviate from the approved Standard Technical Specifications (STSs) for Westinghouse plants (NUREG-1431, Revision 1) with regard to removing items from the TSs and relocating them to the Core Operating Limits Report (COLR). Provide justification for this j deviation from the approved STSs for the quadrant power tilt ratio l (QPTR) and the f 61) and f 2(a!) inputs to the OTAT and OPAI reactor i

protection system,(, considering the Nuclear Steam Supply System (NSSS) and nuclear instrumentation remain Westinghouse designs. Additionally, justify the use of a QPTR limit of 1.03 which is less conservative than the formerly used 1.02.

3. Provide additional basis for not including uncertainties in the F",nM and F",(XY) in the footnotes of TS SR 4.2.2.2 and 4.2.3.2.

4. Why is F",(XYZ) not reduced by 2% over what is specified in the COLR as the approved BAW-10163 prescribes in TS 4.2.2.2.C.4.e.l?

5. Explain why F and F verified with,the in, are corenot verified detectors as each time the the approved excore QPTR BAW-10163 is methodology ,

prescribes in TS SR 4.2.3.3 and 4.2.2.2.c.4.e. l

6. Justify the assumption that 15% of the steam generator tubes have been plugged. What effect will that have on the LOCA analysis results if ,

more or less tubes are plugged in the broken or unbroken loops? (see i

p. 5-6) Additionally, the non-LOCA analysis assumes that 20% of the l tubes are plugged. Justify why this is limiting for all transients  !

analyzed. l

7. The loss-of-coolant accident (LOCA) analysis assumes that the reactor coolant system (RCS) flow is 348,000 gpm; however, the TS Figure 3.2-1 allows flow down to 342,000 gpm if power is derated. Show that the deration is sufficient to assure that no limits are exceeded.

Additionally, the TS minimum RCS flow is being reduced with this  !

submittal and the analysis on the new Framatome fuel is performed using the lower flowrate. However, TVA is relying on the current Westinghouse analysis to show no limits are exceeoed for the Westinghouse fuel ,

inserts. Justify the use of the current analysis when the TS minimum  !

f flowrate is going to be lower than was assumed in this analysis.

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Additionally, in section 7.3.2, what is the basis for the equation reducing F, with reactor power?

j 8. Please describe the changes made to the approved Babcock & Wilcox j Nuclear Technologies (BWNT) recirculating steam generator (RSG) a evaluation model (described in p. 5-79) in greater detail and discuss any implications on the prior staff review and approval.

9. Has the core down flow bypass in the baffle region been explicitly

modeled in the LOCA and non-LOCA analysis? Describe how it is modeled.

10. With regard to the fuel design features changed from the approved topical report BAW-10172, describe the changes in greater detail. The submittal is unclear with regard to the bottom nozzle changes. Have

{ these changes been approved by the staff? Describe what testing and ,

reanalysis has been performed on assemblies with the described changes  !

l (structural, flow, CHF). Please verify that the structural analysis performed in chapter 8 of the topical report includes the changes i identified.  ;

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11. The pressurizer heaters and sprays are not modeled for the non-LOCA analysis. The results of some transients are worse if these control features function (i.e., the peak steam generator pressure can be higher

{ if the sprays act to delay a reactor trip on high RCS pressure). The 1 staff safety evaluation (SE) on the methodology requires consideration j of these control features. Describe why these control features are not j modeled.

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12. Discuss in greater detail the implications of no longer modeling a " hot ,

channel" and an " average channel" for non-LOCA transient methodology 1 (described on p. 6-4).

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13. The staff SE for BAW-10169 states that the acceptance criteria for a ,

locked RCP rotor is the 95/95 DNBR criteria; however, the analysis l i predicts DNB and applies the acceptance criteria for infrequent t

incidents (Condition III). Please correct this departure from the 1 approved methodology.  !

14. The acceptance criteria established (pp. 2-3 and 7-1) for events of i moderate frequency (condition I and II) includes a 99.9% probability

} that "DN8 will not occur core wide." The SRP acceptance criteria l I require "at least 99.9% of fuel rods in the core will not experience )

] DNB" (SRP 4.4-3). The two acceptance criteria are not equivalent.  !

! Please correct or clarify the difference.

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15. It is unclear from the submittal which of the chapter 15 analyses were j redone to support the fuel change and which ones were re-evaluated.

i Table 6.1-1 is not consistent with the verbiage in text on a number of

! examples. Table 6.1-1 only lists six transients that were reanalyzed; j however, analysis rerelts are discussed for other transients (for j example, misaligned RCCA discusses analysis results in Section 6.2.3).

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To clarify the situation, state for each Chapter 15 transient whether it was reanalyzed using the Framatome methods with acceptable results or why reanalysis is not necessary (why current analysis remains bounding with the new fuel and a lower RCS flowrate). Include both the effects of the new fuel and the lower RCS flow.

16. There is no justification that a complete loss of flow is more limiting than a partial loss of flow as stated in the submittal. Please provide the justification.

17. For a number of transients (loss of flow), a delayed neutron fraction corresponding to end of life (E0L) is chosen. Describe why this is I conservative.

l 18. Describe how the stuck rod is modeled for each transient and identify which analyses assume a stuck rod and which ones do not.

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19. The transient analysis presented does not discuss single failure l assumptions in any detail. For each transient, state the limiting j single failure chosen and why it is limiting.

20. For the main steam line break analysis (p. 6-54), the volume of water between the cold leg piping and the first check valve is considered to ,

l be at 0 ppm boron concentration. The current Final Safety Analysis '

Report (FSAR) analysis assumes the volume of water in the piping from the RWST to the cold leg piping (a much bigger water volume) is all  :

at 0 ppm. Please describe how you can assure that the water in the ,

piping between the RWST and the first check valve is at least 1950 ppm.  !

21. Explain why there is no flow to the intact steam generators after <

l 20 seconds into the main steam line break transient analysis (Figure l 6.4-13). Shouldn't emergency feedwater start injecting when main feedwater is isolated?

22. A statistical core design (SCD) methodology is used to analyze some of i the transients and used to derive some safety limits, peaking limits,  !

and departure from nucleate boiling ratio (DNBR) limits. Describe how  !

the non-SCD transients are used to provide input to the same limits.

Also, identify the conditions for their application to this reload.

23. Is rod bow explicitly accounted for in the DNBR methodology? If so, where is it accounted for (retained margin, DNBR penalty, peaking factor adjustments)?

24. Why is the RCS pressure used for DNBR purposes (Table 7.1-2) chosen to be 2280 psia when the nominal RCS pressure is 2235 psig?

25. Please provide a justification for the Sequoyah-specific uncertainties chosen to calculate the statistical design limit.

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26. The critical heat flux correlation used for non-LOCA applications is r approved for both the Framatome and Westinghouse fuel; however, it has l

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not been approved for mixed core applications. Provide an appropriate penalty to the limiting rod that will bound the misapplication of the .

critical heat flux correlation. Additionally, the analysis provided  !

does not include the Westinghouse standard fuel in the mixed core penalty. Provide a justification why this is not accounted for.

27. Describe background information and' the bases for those studies related  !

to the Trojan Plant which concluded that a 3% transition core DNBR penalty should be applied to the Mark-BW when it is being inserted into a Westinghouse standard core with respect to the hydraulic compatibility j of the Mark-BW fuel design with the Westinghouse standard design.

Identify the similarity or difference in relation to the transition core DNBR penalty between the Sequoyah and Trojan reloads.

l 28. Provide the final conservatively bounding mixed core configuration for l the SQN mixed core DNBR analysis and the transition penalty based on i assuming that the center hot assembly is either a single Mark-BW fuel assembly in a core of the VANTAGE 5H or a single VANTAGE SH in a Mark-BW core. Also, provide the result of the DNBR analysis using plant- and

! cycle-specific core loading configuration and the same limiting power i

distribution input in the above analyses. Show that the VANTAGE 5H to Mark-BW design peak difference will offset any transition core effects on the VANTAGE 5H and provide the description of the retained thermal l margin in relation to the transition core penalty.

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29. Provide clarification of the limited use of Westinghouse standard reinserts in a SQN transition core application and provide justification that the transition penalty will bound the SQN application.

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30. Provide the bases for obtaining 2% of an increase in liit force for the limiting transition core configuration (one VANTAGE 5H in a Mark-BW core). Also, provide the data for lateral crossflow velocities for the mixed core configuration and an acceptable criterion for lateral crossflow.

31. Form loss coefficients for the fuel subcomponents were determined using the measured pressure drops. A LYNXT hydraulic model using those form l loss coefficients showed that the total pressure drop of the ('

Westinghouse VANTAGE 5H design is approximately 4% higher than that of

, the Mark-BW and that the Westinghouse standard fuel assembly is approximately 5.5% lower in the pressure drop than the current Mark-BW. I Provide the detailed analysis with respect to the overall impact on the mixed core DNBR analysis based on these 4% and 5.5% pressure drops. I Also, describe how the Figure 3.2 is generated and its application to the mixed core DNBR calculation if flow is much greater than 383,000 gpm.

32. The horizontal seismic and LOCA structural loads were calculated for the mixed core for Mark-BW f uel and Westinghouse Standard fuel. Why was Westinghouse V-5H not used?

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in the reactor and describe how the results are affected by the differences.

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34. The nethodology approved for the mixed core structural analysis is contained in BAW-10133. There is no reference to this methodology in the submittal. Please verify that this methodology was used.

35. On p. 8-7, the stated design criteria (with a reference to the Standard Review Plan [SRP]) for the LOCA combined with the SSE does not include l control rod insertability; however, the SRP does require control rod

! insertability for this event. Correct the criteria and verify that control rod insertability is maintained for the combined loads.

36. The submittal states that the target burnups for SQN are 62,000 MWD /mtU for the peak rdd; however, the safety evaluation for Mark-BW fuel only approves the fuel up to burnups of 60,000 MWD /mtU for the peak rod.

Verify that the 3eak burnups will not exceed approved values and i describe how eac1 of the other limitations contained in the Safety Evaluation for:BAW-10172 (Section 6.0 Conclusions) are met, f

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.- Mr. Oliver D. Kingsley, Jr. SEQU0YAH NUCLEAR PLANT

!. Tennessee Valley Authority

]- cc:

Mr. O. J. Zeringue, Sr. Vice President Mr. J. T. Herron, Plant Manager i Nuclear Operations Sequoyah Nuclear Plant

Tennessee Valley Authority Tennessee Valley Authority 1 6A Lookout Place P.O. Box 2000 i 1101 Market Street Soddy Daisy, TN 37379

Chattanooga, TN 37402-2801 Regional Administrator 5

Mr. Mark 0. Medford, Vice President U.S. Nuclear Regulatory Commission j Engineering & Technical Services Region II 4

Tennessee Valley Authority 101 Marietta Street, NW., Suite 2900 6A Lookout Place Atlanta, GA 30323 i

1101 Market Street

Chattanooga, TN 37402-2801 Mr. Melvin C. Shannon 1 Senior Resident Inspector l Mr. R. J. Adney, Site Vice President Sequoyah Nuclear Plant

! Sequoyah Nuclear Plant U.S. Nuclear Regulatory Commission i Tennessee Valley Authority 2600 Igou Ferry Road

P.O. Box 2000 Soddy Daisy, TN 37379 i Soddy Daisy, TN 37379 i Mr. Michael H. Mobley, Director i General Counsel Division of Radiological Health u Tennessee Valley Authority 3rd Floor, L and C Annex

) ET 10H 401 Church Street i 400 West Summit Hill Drive Nashville, TN 37243-1532 j Knoxville, TN 37902 j County Executive i Mr. Raul R. Baron, General Manager Hamilton County Courthouse

Nuclear Assurance and Licensing Chattanooga, TN 37402-2801

! Tennessee Valley Authority i 4J Blue Ridge 1101 Market Street i Chattanooga, TN 37402-2801 I l

? Mr. Pedro Salas, Manager l

Licensing and Industry Affairs Tennessee Valley Authority 4J Blue Ridge 1101 Market Street l Chattanooga, TN 37402-2801 4

Mr. Ralph H. Shell, Manager 4 Licensing and Industry Affairs i Sequoyah Nuclear Plant Tennessee Valley Authority l P.O. Box 2000

)

Soddy Daisy, TN 37379 I

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