ML19327B490

Responds to NRC 891003 & 04 Requests for Addl Info Re Proposed Vantage 5 Tech Spec Amend.Demonstration Program to Determine Early Performance Data on Vantage 5 Fuel Assembly Design Features Successfully Performed

ML19327B490
Person / Time
Site:	Byron
Issue date:	10/19/1989
From:	Chrzanowski R COMMONWEALTH EDISON CO.
To:	Murley T Office of Nuclear Reactor Regulation
References
NUDOCS 8910310290
Download: ML19327B490 (6)

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/ ommonwealth Edison 72 West Adams Street, Chicago, Illinc!s i NT X6dissleply to: Posi OllE B~oT767~  !

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' ]n Chicago, Illinois 60600 0767 i J

s, October 19, 1989 ') 1 7, ,.

l Dr. Th'omas E. Murley, Director Office of Nuclear Reactor Regulation .

'U.S. Nuclear Regulatory Commission ,

F . Washington, DC 20555  ;

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Subject:

Byron Station. Units 1 and 2 p Proposed Technical Specification Amendment

p. - Response to Request for: Additional Information '

l Operating Licenses NPF-37 and NPF-66 NRC-Docket Nos. 50-454 and 50-455 '

Reference:

(a) July 31, 1989, letter from R. A. Chrzanowski  !

to T. E. Murley  ;

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Dear Dr. Murley:

A request to amend the Byron Station Units 1 and 2 Technical Specifications was submitted in reference (a). Through telephonic conversations held on October 3 and 4, 1989, between Commonwealth Edison and h members of your staff, questions arose regarding the proposed Vantage 5 .

amendment. The Attachment-to this letter provides the response to those questions. -

Please direct an further questions regarding this topic to this office. .j L Very truly yours, r L

l OS$u l

R. A Chrzano ski Nuclear Li ensing Administrator Attachment l'.

Byron Resident Inspector Ac9l '

cc: ' '

L. N. 01shan - NRR S. Sun - NRR Region III Office s Office of Nuclear Facility Safety 8910310290 891019 PDR ADOCK 05000454 P PDC

1 BESPONSES_TO_NRG.STAEEQUES110NS_ON IHE PROPOSED

,, ** USELOENAhrIAGEJ.EUELAT_THERYRONSBNDWOOD_SIAllONS_UNES 1 ANDJ m ,

NBC-lmposeCLLlmitations on Une of WCAP-10444 The NRC Staff reviewed Westinghouse's WCAP-10444, " Reference Core Report VANTAGE -

5 Fuel Assembly," and concluded in a staff Safety Evaluation Report (SER), Reference 1, that the generic topical report was an acceptable reference to support plant specific ,

applications for use of VANTAGE 5 fuel, provided thirteen conditions identified in the SER were addressed by the licensees. These thirteen conditions were appropriately considered in .

Commonwealth Edison's submittal for License Amendment Request, Reference 2. Each of I the thirteen conditions is addressed below, and a reference is provided for the specific .

section in the Commonwealth Edison Licensing Amendment Request or other appropriate documentation where the condition is discussed.

I NRC Condition.t; The statistical convolution method described in WCAP-10125 lbr the evaluation of initial fuel rod to nozzle growth has not been approved. This method should not be used in VANTAGE 5.

Re.sponsE The statistical convolution method was not used for fuel evaluation. To determine the initial fuel rod-to-nozzle growth gap from fuel rod irradiation growth, the worse-case fabrication +

tolerances were used to evaluate fuel rod performance as discussed in Section 2.0, Page 8, .

of Attachment 1 to Reference 2. This evaluation was in compliance with Condition 1. Fuel I rod performance for all fuel rod designs was shown to satisfy the Standard Review Plan fuel rod design bases on a region by-region basis.

NRC Condition 2:

For each plant application, it must be demonstrated that the LOCA/selsmic loads considered in WCAP-9401 bound the plant In question; otherwise, additional analysis will be required to demonstrate the fuel assembly structuralIntegrity.

B99900891 The LOCA/ seismic loads considered in WCAP-9401 bound the Byron /Braldwood Stations '

Units 1 and 2. This is addressed in Section 2.0, Page 13, of Attachment 1 to Reference 2.

l NBCSondilian2 An irradiation demonstration program should be performed to provide early confirmation performance data for the VANTAGE 5 design.

BRSpoDSel A demonstration program was successfully performed to determine early performance data on the VANTAGE S fuel assembly design features. The VANTAGE 5 demonstration program l at commercial reactors is described in Section 1.0, Page 3 of Attachment 1 to Reference 2.

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NBCLcondition 4 For those plants using the ITDP, the restrictions enumerated in Section 4.1 of this report .

. (SER) must be addressed and infomtation regarding measurement uncertainties must be  !

'provided.

Bespanael '

l Westinghouse has addressed the restrictions enumerated in Section 4.1 of the NRC's i generic VANTAGE 5'SER in a Westinghouse letter to the NRC Staff in Mr.rch 1985,  ;

Reference 3. The ITDP instrument uncertainty methodology used for the Byron /Braldwood  !

Units with VANTAGE 5 fuel was presented in WCAP-11656, Reference 4. Measurement ,

uncertainties were provided to the NRC Staff in Commonwealth Edison's letter, Reference 5.

'NRC Condition 5; The WRB-2 correlation with a DN8R limit of 1.17 is acceptable for appilcation to 17x17 VANTAGE 5 fuel. Additional data and analysis are required when applied to 14x14 or 15x15 fuel with an appropriate DNBR Ilmit. The appIIcability range of WRB-2 is specified in Section 4.2.

Besponsa; VANTAGE 517x17 fuelis proposed to be used at the Byron /Braldwood Stations Units 1 and i

2. As described in Section 4.0, Page 17 of Attachment 1 to Reference 2, the WRB-2 l correlation with a DNBR limit of 1.17 was used for the VANTAGE 5 fuel with the t NRC-approved ITDP methodology. The WRB 2 correlation is supmrted by the DNB test  !

data contained in Apandix A to WCAP-10444-P-A, and was appled within its approved range of applicability "or the Byron /Braidwood Units.

NRC Conditiaah For 14x14 and 15x15 VANTAGE 5 fuel designs, separate analyses will be required to determine a transitional mixed core penalty. The mixed core penalty and plant-specific safety ma In to compensate for the penalty should be addressed in the plant Technical ,

tion Bases.

Besponnel As noted above in Condition 5, VANTAGE 517x17 fuel is proposed for the Byron /Braldwood Station Units 1 and 2. The Westinghouse transition core DNB methodology as applied to the Byron /Braidwood Units is discussed in Section 4.0 of Attachment 1 to Reference 2. The Commonwealth Edison submittal, Reference 2 incorporates the NRC-approved change to the generic VANTAGE 5 transition core effects, Reference 6. The transition core penalty is covered by the margin maintained between the design and safety analysis DNBR limits. The proposed changes (see Attachment 2 to Reference 2) to Technical Specification Section 2.1.1, Reactor Core Safuty Limits Bases, address these margins.

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1 NRC Coadtloah 1

Plant specific analysis should be performed to show that the DNBR limit will not be violated with the higher value of FAH.

. Besp00sE The core DNB methodology as applied to the Byron /Braidwood Stations Units 1 and 2 with VANTAGE 5 fuel is presented in Section 4.0, Page 17, of Attachment 1 to Reference 2. The Commonwealth Edison submittal, Reference 2, contains Byron /Braidwood plant-specific analysis results in Section 5.0 of Attachment 1 which support the use of FAH of 1.65 durin;l ,

the transition period and with a full core of VANTAGE 5 fuel. All safety criteria are met witi j an FAH of 1.65 as demonstrated in Section 5.0. j NBC_Gondition 8:

The plant-specific safety analysis for the steam system piping failure event should be ,

performed with the assumption ofloss of offsite power if that is the most conservative case. 1 Besponsm The Byron /Braidwood plant specific analysis did not require a new analysis for the Main Steam Piping Rupture event. The current analysis is the Byron /Braidwood UFSAR which included both the with and without offsite power cases is still applicable. The DNB design l basis was confirmed for the fuei transition by evaluation as presented in Section 15.2.4 of -

Appendix B of Reference 2.

NBC Goodtionk With regard to thee RCS pump shaft selsure accident, the fuel failure criterion should be the 95/96' ONBR limit. The mechanistic method mentioned in WCAP-10444 is not acceptable.

Besponas The mechanistic method was not used for the RCS Sump shaft selsure (locked rotor) accident addressed in Section 15.3.3 of Attachment D to Reference 2. Any rods which violated the 95/95 DNBR limit were assumed to fall. Westinghouse performed two separate and distinct analyses for the locked rotor / shaft break event. The first analysis, using the LOFTRAN and FACTRAN computer codes, was conducted to determine the 3eak RCS pressure, the peak clad temperature, and the amount of zinc-water reaction. T1ese results are presented in Table 15.3.2 of the VANTAGE 5 RTSR and were not used in any way to determine the number of rods in DNB.

A second analysis, using the LOFTRAN, FACTRAN, and THINC computer codes, was performed to determine the number of rods that experience DNB during the accident. The results of this analysis, which included increase in the Fo and FAH, showed that 0% of the .

rods were predicted to be below the 95/95 DNBR limit. To reiterate, the mechanistic approach identified in Reference 9 (NUREG-0562) of WCAP-10444 was no.t used in the ~

Byron /Braldwood VANTAGE 5 License Amendment Request.

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NRC_Gondition 10: ,

, .. lf a positive MTC is intended fcr VANTAGE 5, the same positive MTC consistent with the plant Technical Specification should be used in the plant specific safety analysis. .

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- Response;

. Commonwealth Edison does not have any current plans to incorporate a positive MTC in the .

operation of the Byron /Braidwood Stations Units 1 and 2. Therefore, this condition is not  !'

currently applicable.

NBC.Gaadition.LI: ,

The LOCA analysis performed for the reference plant with higher Faof 2.55 has shown that ,

the PCT limits of 2200*F is violated during transitional mixed core, Plant specific LOCA analysis must be done to show that with the appropriate value of Fa the 2200*F criteria can be met during use of transitional mixed core.

Besponse;

l. In accordance with Condition 11 of the VANTAGE 5 NRC SER, Byron /Braidwood Station Units 1 and 2 specific LOCA analyses were performed with consideration of transitional core

- effects. The Large Break LOCA analysis is summarized in Section 5.2.1 of Attachment 1 to Reference 2, and detailed results are also provided in Section 15.6.5 of Attachment 4 to

,- Reference 2. As described therein, the ECCS acceptance criteria of 2200*F is met for the i'

! Byron /Braidwood Station with a LOCA Fo o f 2.50. The worst case n i alad temperature is

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1933.1 F;it includes a conservative transltion core penalty of 50 F.

NBC_Gooditico 12:

Our SER on Westinghouse's extended bum topical report WCAP-10125 is not yet complete; the approval of the VANTAGE 5 de n for operation to extended bumop levels is contingent on NRC approval of WCAP-10125. wever, VANTAGE S fuel may be used to those burnups to which Westinghouse fuel is presently operating. Our review of the Westinghouse extended burnup topical report has not identified any safety Issues with '

operation to the burnup value given in the extended bumup report.

heBp0D8R1 WCAP-10125 has been approved (see Reference 7). The extended burnup methodology contained in this topical has been applied and is addressed in Section 2.0, page 8, of Attachrnent 1 to Reference 2.

NBG_Gondifiaa.13:

Recently a vibration problem has been reported in a French reactor having 14 foot fuel assemblies; vibration below the fuel assemblies in the lower portion of the reactor vesselis damaging the movable incore instrumentation probe thimbles. The staff is currently evaluating the implications of this problem to other cores having 14 foot long fuel bundle assemblies. Any limitations to the 14-foot core design resulting from the staff evaluation must be addressed in plant specific evaluations.

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Beaponte; r The Byroneraidwood Stations Units 1 and 2 have 12 foot long fuel assembly bundles and 4 therefore the above condition is not applicable.

Reference:

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1. to Westinghouse (E.P. Rabe) Re: 'AccoMance for NRC Staff of Referencing letter (C.O.Topica Licensing Thomas)l Report, WCAP 10444, VANTAGE 5 Fue Assembly,'

undated.

2. NRC), Re: " Application for Letter Amendmentfrom toRAFacility Chrzanowski 0:Merating (CECO)

LicensestoNPF T.E.37Murleyan (d NPA77. NRC Docket Nos.

50 454 and 50-455, detec July 31,1989 and letter from S.C. Hunsador (CECO) to T.E.

Murley (NRC), Re: ApMication for Amendment to Facility Operating Licenses NPF 72 and NPF-77, NRC Doc tet Nos. 50-456 and 50 4b7, dated October 19,1989.

3. Westinghouse letter, E.P. Rahe Jr. to C.O. Thomas (NRC), Response to Roquest No.1 for adetional Information on WCAP 10444 entitled ' VANTAGE 5 Fuel Assembly' (Proprietary) NS NRC_85-3014, dated March 1,1985.

4. " Westinghouse improved Thermal Design Procedure Instrument Uncertainty Methodology,' WCAP 11656, December 1987.

5. to H.R..Denton (NRC) Byron /Braidwood Stations Units 1 Letter from Docket and 2 ITDP," T. Tramm Nos. (CECO)454,50-455,50-456 50- and 50-457, May 1982.

6. Westinghouse letter, W.J. Johnson to M.W. Hodges (NRC), NS NRC-87 3208,

" VANTAGE 5 DNB Transition Core Effects," October 2,196.

7. Extendec. Burnup Evaluation of Westinghouse Fuel" WCAP 10125 P A, December 1985.

NRO_Concera inconsistency of values in FAH (B21 and B2 2) and 3/4 2 8.

L Besponse:

The values in both Sections of the Technical Specifications are consistent. The difference l can be explained by the application of a 4% measurement uncertainty applied to the factor 1.49 factor in the bases Section resulting in a f actor of 1.55 in Section 3/4 2-8.

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