ML20081D129

Forwards Proprietary Info to Support NRC Review of Proposed Amends to Licenses NPF-9 & NPF-17.Info Changes Operating Limitations Affected by Optimized Fuel Assembly Design for Unit 1/Cycle 2 Reload.Info Withheld (Ref 10CFR2.790)

ML20081D129
Person / Time
Site:	Mcguire, McGuire
Issue date:	03/09/1984
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML19292D029	List: ML20081D129
References
NUDOCS 8403150156
Download: ML20081D129 (5)

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g. 4 DUKE POWER COMI%NY P.O. Box 03180 C11AnLOTW, N.C. 28242 l

IIALH. TUCKER Teterneoxe '

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WEELEAR PeoptmO4 (704) 373-4531 March 9, 1984 N$g ~I r

Mr. Harold R. Denton, Director It Office of Nuclear Reactor Regulation 9 U. S. Nuclear Regulatory Commission 4 [?[h,'!/,'y .

Washington, D. C. 20555 '#!jf Attention: Ms. E. G. Adensam, Chief

/ Licensing Branch No. 4 '

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370 McGuire 1/ Cycle 2 OFA Reload

Dear Mr. Denton:

My letter of December 12, 1983 transmitted Proposed License Amendments to Facility Operating Licenses NPF-9 and NPF-17 for McGuire Nuclear Station Units 1 and 2, respectively. This submittal, which basically changes plant operating limitations given in the Technical Specifications affected by use of the Optimized Fuel Assembly (OFA) Design for McGuire Unit 1/ Cycle 2 to ensure plant operation consistent with the Design and Safety Evaluations, was subsequently revised by my letter dated February 20, 1984.

Ms. E. G. Adensam's letter dated March 5, 1984 requested additional information necessary to support NRC Staff Review of the Proposed Amendments.

Accordingly, this information is provided in the attachment to this letter.

Should there be any further questions concerning this matter, please advise.

Please note that this submittal contains information proprietary to Westinghouse Electric Corporation. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations. An Application for Withholding will be submitted later along I

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e Mr. Harold R. Denton Attention: Ms. E. G. Adensam March 9, 1984 Page 2

' with an affidavit signed by Westinghouse, the owners of the information, which will set forth the basis on which the information may be withheld from public disclosure by the Commission.

Very truly yours, ,

al B. Tucker PBN:je Attachment cc: Mr. J. P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission-Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 Mr. Dayne Brown, Chief Radiation Protectiou Branch Division of Facility Services Department of Human Resources P. O. Box 12200 Raleigh, North Carolina 27605 -

Mr. W. T. Orders Senior Resident Inspector McGuire Nuclear Station i

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DUKF POWER COMPANY MCGUIRE NUCLEAR STATION MCGUIRE 1/ CYCLE 2 0FA RELOAD - ADDITIONAL INFORMATION

1. Justify the use of the W-3 DNBR correlation for the steam line break event and show that all the parameters are within the range of j applicability.

RESPONSE: The W-3 DNBR correlation is used for the steamline break event since the pressure falls outside the range of applicability for the .

WRB-1 DNBR correlation (i.e., 1440 $ P 5 2490 psia). Moct references give the pressure range for the W-3 DNBR correlation as greater than 1000 psic (Reference 1). However, evaluations using the same source of data as used in the development of the W-3 correlation have shown that the pressure range can be extended below 1000 psia. As shown in the attached figure (No. 1), no abnormality exists for the low pressure data. This figure was taken from Reference 2, and was used to license the extension of the W-3 correlation to low pressures for Prairie Island.

The parameters of the limiting steamline break case for McGuire 1/ Cycle 2 are as follows:

Pressure

• 1200 psia Mass velocity N 2.2 x 106 lbm/hr-ft 2 Local quality N 5.9%

All of these parameters fall within the range of applicability of the W-3 DNBR correlation.

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2. Justify the use of 2% mixed core penalty ins";ead of 5%.

RESPONSE: The transition core DNB methodology given in Reference 3 is applicable for McGuire's transition from 17x17 standard fuel to 17 x 17 optimized fuel. A comprehensive review of all transition patterns versus an all 0FA pattern over appropriate ranges indicate that a two percent DNBR penalty is sufficient to bound these analyses. The actual value cf -

1.9% DNBR is given in Table 2 of Reference 4. The value of five percent i

used for WCAP-9500's generic application is unnecessarily conservative.

i 3. For transients analyzed to determine fuel failure provide the DNBR value as a function of time.

RESPONSE: ' Duke Power Company int erprets this to mean Condition II a

Transients for which DNB is a uriteria/ concern. Mr. H.LB. Tucker's-

, December 12, 1983.McGuire 1/ Cycle 2 0FA Reload License Amendment Submittal' (Attachment 2A) included detailed Non-LOCA Accident Analyses of the~

McGuire Units 1 and 2 FSAR impacted by the proposed changes. Included in the analyses are the following figures.which provide .the requested information on DNBR versus time: Figures 15.1.1-2, 15.1.2-2, 15.1.2-4, 15~.1.2-6,15.1.2-8,15.2.3-2,15.2.3-4,'15.2.3-6,Ifl.2.3-8,15.3.1-4,

.15.3.2-4, 15.4.2-3, 15.4.2-6, 15.4.4-5, 15.5-3, and 15.6-1. . In addition,

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Figures 15.4.2-7, 15.4.2-8, and 15.4.2-9 provide DNBR versus insertion rate for appropriate cases.

4. Provide a description of plans for post-irradiation poolside surveillance of 0FA fuel.

RESPONSE: A routine fuel inspectioa program will be implemented on the irradiated and discharged optimized fuel from the initial reload region.

The program will involve visual examinations on a representative sample of assemblies from the initial fuel region at each refueling tntil this fuel is discharged. Visual observations will include, but not be limited to, crud buildup, rod bowing, grid strap conditiens and missing components. Additional fuel inspections would be performed depending on the results of operational monitoring, including coolant activity, and the visual fuel inspections.

5. Provide information on whether appropriate seismic and LOCA forces are bounded by the cases r.onsidered in WCAP-9401 or additional analyses.

RESPONSE: An evaluation of fuel assembly structural integrity considering the lateral effects of a LOCA combined with a seismic accident was evaluated for McGuire I and 2. Results of the evaluation concluded that both units are bounded by the results of the seismic and LOCA analyses detailed in WCAP-9401.

6. Provide supplemental ECCS calculations using NRC supplied LOCA cladding models.

RESPONSE: The calculations performed for McGuire 1/ Cycle 2 have employed UHI evaluation model code versions which include NUREG-0630 BURST / BLOCKAGE models. Therefore, no supplemental calculations are necessa ry. ,

7. Which WCAP was used for the updated version of the non-UHI ECCS evaluation model?

RESPONSE: WCAP-9220-P-A, " Westinghouse ECCS Evaluation Model," Rev. 1, February 1982.

8. What degree of symmetry will the Cycle 2 core have? Were the design calculations for Cycle 2 done on a full core basis? If not, please juatify the type of calculations used in light of the fact that the Cycle 1A loading did not have the same degree of symmetry as the normal first cycle core. What other effects might the reduced symmetry of Cycle 1A have upon future cycles especially Cycle 27 How are these effects being compensated for?

RESPONSE: During Cycle 1 problems encountered with broken burnable poison holddown springs led to the decision to remove all the burnable poison assemblies except the two which also contained the secondary neutron source rods. These assemblies were in core locations H-3 and 4

H-13 along the core major axes (see Figure 2). Because the asymmetry was along the mejor axes, a quarter core model could be used to accurately model the half core symmetric problem.

In setting the fuel loading pattern for Cycle 2, the assemblies in H-3 and H-13 were shuffled to core locations H-7 and H-9 respectively (see Figure 3). Again, since these locations are on the major core axes, a quarter core model could be used to accurately model the half core symmetric problem.

9. What is the assumed percentage of bypass flow? Justify your assumption.

RESPONSE: The thermal design bypass flow for McGuire Unit 1 is 7.5%

which includes rod cluster control guide thimble cooling flow, head cooling flow, cavity flow, baffle leakage, and leakage to the vessel outlet nozzle.

A nominal core bypass flow of 6.0% (consistent with WCAP-9500 UHI plant analyses) is used in the Improved Thermal Design Procedure (ITDP) as explained in Reference 5.

10. In our Safety Evaluation Report on WCAP-9500 " Reference Core Report 17 x 17 Optimized Fuel Assembly", the staff required that those plants using the Westinghouse Improved Thermal Design Procedure (ITDP) supply addition-al information on the plant specific application of the IfDP. Since the applicant is using the ITDP to perform their thermal-hydraulic analyses, l they must comply with the following: I

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(1) Provide the sensitivity factors (S 1) and their range of applicability:

RESPONSE: The sensitivity factors (S g

) and their range of applicability are given in T1ble 1 for McGuire Unit 1. Please note that these values are the same as those used in WCAP-9500 with the exception of the range for Vessel Flow. The range on flow for McGuire Unit I has been extended down to 275520 GPM (70% flow) with no change in the corresponding sensitivity factor being required.

(2) If the S. values used in the McGuire analyses are different than those used in WCAP-9500, then re-evaluate the use of an uncertainty allowance for application of equation 3-2 of WCAP-8567, " Improved Ther.nal Design Procedure" and the linearity assumption must be validated: l RESPONSE: Theg S values used in the McGuire Unit 1 analyses are the same as those used in WCAP-9500. Therefore, re-evaluating the use of an uncertainty allowance for application of equation 3-2 of WCAP-8567, " Improved Thermal Design Procedure" and the linearity assumption is not required.

(3) Provide and justify the variances and distributions for input pa rameters :

RESPONSE: -This information is supplied in Attachment No. 1. Duke Power Company provided Westinghouse with instrumentation -uncertainty data to be used in conjunction with ITDP for the McGuire Nuclear Station. The data provided by Duke was reviewed and determined that it was consistent with, or conservative with respect to, the 1

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assumptions made in the attached docuacnt. Please note that the attached document is Westinghouse proprietary.

(4) Justify that the normal conditions used in the analyses bound all permitted modes of plant c i eration:

RESPONSE: This information is supplied in Attachment No. 1.

(5) Provide a discussion of what uncertainties, including their values, are included in the DNBR analyses:

RESPONSE: The uncertainties included in the ITDP DNBR analyses for McGuire Unit 1 are given in Table 1. As a result of these values being different from those used in WCAP-9500, the Design DNBR Limits also differ. The calculation of t6'e Design Limit DNBR's for the Typical and Thimble cells are given in Tables 2 and 3 respectively.

Please note that Tables 1, 2 and 3 are Westinghouse Proprietary.

(6) Provide a block diagram depicting sensor, processing equipment, computer and readout devices for each parameter channel used in the uncertainty analysis. Within each element of the block diagram identify the accuracy, drift, range, span, operating limits, and setpoints. Identify the overa'll accuracy of each channel transmitter to final output and specify the minimum acceptable accuracy for use with the new procedure. Also identify the overall accuracy of the final output value and maximum accuracy requirements for each input channel for this final output device.

RESPONSE: This it. formation is supplied in Attachment No. 1.

(7) If there are any changes to the THINC-IV correlation, or parameter values outside of previously demonstrated acceptable ranges, the staff requires a re-evaluation of the sensitivity factors and of the use of equation 3-2 of WCAP-8567.

RESPONSE: For McGuire Unit 1, the THINC-IV code and WRB-1 DNB correlation are the same as that used in WCAP-9500. Therefore, re-evaluating the sensitivity factors and the use of equation 3-2 of WCAP-8567 is not required. ,

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References:

1. Tong, L. S., AEC Critical Review Series, " Boiling Crisis and Critical Heat Flux", TID-25887, August 1972.

2. Prairie Island FSAR, Amendment 20, P. 14.2-30, Docket #50-282, August 4, 1972.

3. Davidson, S. L., Iorii, J. A., " Reference Core Report-17x17 Optimized Fuel Assembly", WCAP-9500-A, May 1982.

4. Letter from E. P Rahe to J. R. Miller, NS-EPR-2643, dated August 17, 1982, entitled " Supplement to WCAP-9500 and WCAP-9401/9402 NRC Safety Evaluation Report (SER) Mixed Core Compatibility Items Supplemental Information.

5. Chelemer, H. , Boman, L. H. , and Sha rp, D. R. , " Improved Thormal Design Procedure", WCAP-8567-P, July 1975 (Proprietary) and WCAP-8568, July 1975 (Non-Prop rieta ry) .

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