ML20205Q203

Forwards Addl Info in Support of Application for Amend to License NPF-30 Re Reload

ML20205Q203
Person / Time
Site:	Callaway
Issue date:	02/18/1986
From:	Schukai R UNION ELECTRIC CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML19292F324	List: ML20205P890 ML20205P897 ML20205Q203
References
ULNRC-1258, NUDOCS 8605280207
Download: ML20205Q203 (13)

Text

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p C0ce 3 UNION WCTRIC COMPANY P

Coo *W c" February 18, 1986 8

1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Cc= mission Washington, D. C. 20555

Dear Mr. Denton:

ULNRC-1258 DOCKET NUMBER 50-483 CALLAWAY PLANT RELOAD LICENSE AMENDMENT REQUEST ADDITIONAL INFORM.ATION

References:

1) ULNRC-1207 dated 11/15/85

2) ULNRC-1227 dated 12/13/85

3) ULNRC-1247 dated 1/28/86 The referenced letters transmitted the reload license amendment requ,est for callaway Cycle 2 along with additional information to support this request.

The attachment to this letter documents clarifying information provided in a February 11, 1986 phone call involving NRC, Westinghouse, and Union Electric personnel.

The information provided was available from the previous analysis and as a result no additional analysis was required.

The attachment contains information which is proprietary to the Westinghouse Electric Corporation.

Due to the proprietary nature of the material contained in this attachment, which was obtained at considerable Westinghouse expense and the release of which would seriously affect their competitive position, we request that this information be withheld from public disclosure in accordance with the Rules of Practice, 10 CFR 2.790, and that the information presented therein be safeguarded in accordance with 10 CFR 2.903.

We believe that withholding this information will not adversely affect the public interest.

This information is for your internal use only and should not be released to persons or organizations outside the Directorate of Regulation and the ACRS without prior approval of Westinghouse Electric Corporation.

Should it become necessary to release this information to such persons as part of the review procedure, please contact Westinghouse Electric Corporation and they will make the necessary arrangements required to protect their proprietary interests.

$o00 $DbOk5

$3 P

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.If there are additional questions,"please contact us.

Very truly yours,

(

Rob rt J Schukai Gene Manager - Engineering RJS/bjk Attachments

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1.

• WESTINGNOUSE PROPRIETARY class 3 Item _1 Provide a table, similar to Table 4.4-1 of the SNUPPS FSAR, for-comparison of CTA and LOPAR fuel types.

The table should include a comparison of core pressure drop.

Response

Refer to attached Table 1, Callaway Thermal /Eydraulic Design Parameters.

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l 1

Tuu 1 CALLAWAY THERMAL / HYDRAULIC DESIGN PARAMITERS Cycle 2 Cycle 2 Operating

? Design thermal and Hydraulic Desien Parameters Parameters Parametecs (Using 11DP)

Reactor Core Heat Output, MWt 3411 3565 0

Reactor Core Heat Output, 10 BTU /hr 11,640 12,164 feat Generated in Fuel, t 97.4 97.4

, Core Pressure, Nominal, psia 2280 2280 Minimum DNBR at Neminal Conditions Typical Flow Channel Call

>2. 2 7 (OFA) 2.27 (OFA)

22. 2 7 (LOPAR)

>2.27 (LOPAR)

Thimble (Cold Wall) Flow Channel Cell

>2.14 (OFA) 2.14 (OFA)

>2.14 (LOPAR)

>2.14 (LOPAR)

Dasign DNBR for Design Transients Typical Flew Channel Cell 1.45 1.45 Thimble (Cold Wall) Flow Channel Cell

~1.42 1.42 tR Correlation WRB-1 WRB-1 RTP Wominal Coolant Conditions Vassel Minimum Measured Flow 0

Rate (Including Bypass), 10 lbm/hr 141.9 141.9 GPM 382,630 382,630 2

vossel Thermal Design Flow 6

Rate (Including Bypass), 10 lbm/hr 138.8 138.8 GPM 374,400 374,400 Core Flow Rate (Excluding Eypass, based on TgT),

10 lbm/hr 130.7 130.7 GPM 352,690 352,690 Based on FSAR Safety Analysis Tin = 560.1*F, 2280 psia 2Includes 10% steam generator tubes plugged

r -

WESTINGHOUSE PROPRIETARY CLASS 3 TABLE 1 (CONTINUED)

Cycle 2 Cycle 2 Operating

- Design RTP Nominal Coolant Condi_tions Parameters Jarameters 51.0 8 (LOPAR) 3.'54.13 54.13(OTA)3 (OTA)3 Fuel Assembly Floy Area for Heat Transfer, ft 51.08 (LOPAR) 6 2

core Inlet Mass velocity, 10 lbm/hr-ft 2.41(OTA) 2.41 (CTA)3 (Based on TDF) 2.56(LOPAR)3 2.56(LOPAR) l l

l 3Assures all OFA or all LOPAR Cere

WESTINGMOUSE PROPRIETARY CLASS 3 TABLE 1 (CONTINUID)

Cycle 2 Cycle 2 n/P Thermal and Hydraulic Desien Parameters Operating Design (Based en Thermal Design Flow)

Parameters Parameters 2,

Nctinal Vessel / Core Inlet 558.2 556.8 Temperature,

'T vossel Average Temp., 'F 588.5 588.4 l

Cere Average Temp.,

'T 591.9 592.0 Vossel Outlet Temp., 'T 618.8 620.0 Average Temperature Rise 60.6 63.2 in vessel,

• F Average Temperature Rist 63.9 66.6 in Core,

'T Neat Transfer Active Beat Tgansfer Surface 57,505(OFA) 57,505 (OFA)3 197,200(OFA)3)3 (CFA)g)3 59,742 (LOPA 59,742 (LOPA Area, ft 2

206,090 Average Esat Flux, BTU /hr-ft 189,800(LOPAR)3 198,370 (LOPAR)3 I

average Linear Power, kw/ft 5.44 5.69 Poak Linear Poyer for Normal 12.63 13.20 Operation, kw/ft Tcmperature at Peak Linear Power 4700 4700-for Prevention of Centerline

Malt,

'T Pressure Drop' Across' Core, psia 26.4 1 2.6 (ora)3 26.4 3 2.6 (OTA) 26.5 1 2.6 (LOPAR)3 26'.5 1 2.6 (LOPAR) 3 2Based on FSAR Safety Analysis Tin = 560.1,,'F, 2280 psia 3Assumes all 0FA or all LOPAR core 4 Based on 2.32 T Peaking Factor g

5Based on bes: estimate reactor flow rate

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r-WESTINGROUSE PROPRIETARY CLASS 3 Itom 2 Provide a description of the core bypass flow paths (thimble tube cooling) and the impact of WABA's on thimble tube cooling.

?

Res_eonse:

Corebypassflowisdefinedasthetotalamountofreact5r coolant flow which bypasses the core region.

The following flew paths are identified as bypass:

1)

Baffle-Barrel Region 2)

Head Cooling spray Nozzles 3)

Fuel Assembly / Baffle Plate Cavity 4)

Outlet Nozzle (flow which goes from the downcomer directly to the outlet nozzle) 5)

Thimble Tubes Thimble tube bypass is the flow which cools core component rods.

Although this flow is partially effective in cooling the core, it is not considered as such and is treated as a bypass flow. [

+a,c

] percent of the total bypass flow is allocated for thimble tube cooling in the Callaway Plant.

The total thimble tube bypass flow is the sum of the following individual component bypass flows:

Thimble Tube with

- thimble plug

- control rod

- source rod

- burnable absorber Instrument tube without flux detector For the callaway cycle 2 core,' Wet Annular Burnable Absorbers (WABAs) will be used instead of the glass absorbers of the cycle 1 core.

Since the WABAs provide an additional bypass flow path (in the annulus of the absorber),. they will slightly increase the total thimble tube bypass flow.

Reference 1 provides an upper l

limit to the number of WABA rods required for various core configurations.

For Callaway, this upper limit is 1600 rods while the Cycle 2 design will utilize approximately 1100 WABA rods.-

The total thimble tube bypass flow for Callaway Cycle 2

+a,c (including the WABA rods) is calculated to be within the[]L limit.

Item 3 Provide a summary description of how the.RCS flow measurement uncertainties are treated.

Specifically; address uncertainty attributed to feedwater venturi fouling.

l

WESTINGHOUSE PROPRIETARY CLASS 3 e

R'es-ense Appendices A and.3 of Reference 2 contain the informatio necessary to support the value for RCS flow measurement' uncertainty of 12.2% used in the Callaway Technical specifications.

Appendix A centains a discussion of the generic Westinghouse methodology for calculation of uncertainties used in the Improved Thermal Design Procedure process.

The values referenced in tables therein are typical values provided for example purposes and are not applicable to Callaway.

Plant specific calculations were performed for Callaway and are documented in Appendix 3.

Tables 6, 7 and 8 of Appendix 5 contain the information needed to calculate the uncertainty in the precision flow calorimetric performed once at the start of a cycle.

Table 6 presents the uncertaintier associated with the actual instrumentation at Callaway used for the precision flow calorimetric.

Table 7 reflects the sensitivities used in calculating the actual flow uncertainity.

Table 8 depicts the calculation of the RCS prqcision flow calorimetric senstrament uncertainty.

This value

+a'c of

) flow includes a bias of

4. flow to account for the thirmaI nonrepeatability of the 31rtin pressurizer pressure transmitters.

Since it is impracticable to perform a precision flow calorimetric every time it is needed to verify flow per i

Technical Specification requirements, callaway normalises the elbow taps and calculates a RCS flow directly from these instruments using information from the precision calorimetric.

Therefore, the accounting of instrument uncertainties is I

required.

Table 9 provides the instrument uncertainties and l

ccmbines them with the precision flow calorimetric uncertainties

]t bias) to calculate a total flow uncertainity +a,c

[

plus a{ THe value of +2.2% in the proposed Technical sf+}2.1%flow.

Specification is the result o7 adding 0.1% flow to the flow uncertainity above to account for feedwater venturi fouling.

Itam 4 Provide justification for use of the W-3 DNBR correlation at the low pressure (<1000 psia) associated with the Callaway steamline break transient.

Response

Reference 3 describes the results of applying the W-3 correlation over its original pressure range, 1000 to 2300 psia.

The mean measured-to-predicted critical heat flux ratio and sample j

standard deviation from that analysis are 'shown in Table 2.

Low pressure (500-1000 psia) data were taken from the same sources as those used in the development of the W-3 correlation and were analyzed using the M-3 correlation.

As shown in rigure 1, the results show no anomalous behavior for the low pressure data.

= _.

3 WESTINGHOUSE PROPRIETARY CLASS The W-3 correlation statistics have been recalculated for the (P=500-2300 psia).

The revised statistics are extended database (Table 2) / The essentially unchanged from the original values limit DNBR was also recalculated using the method of Owen-As show statistics demonstrate that there is a 954 probability with 954 (Reference 4).

confidence that DNB will not occur if the minimum DNBR isAgain, this maintained in excess of 1.33.

unchanged from the limit DNBR associated with the original database.

A small number of Westinghouse red bundle test data points with conditions representative of steamline break are available.

3.

Figure 2 shows the rod These were also analyzed ttsing W data of Figure 1.

The bundle data added to the in-tube measured-to-predicted critical heat flux ratios show good agreement with the in-tube data, This evaluation damenstrates that the pressure range of the W-3 correlation can be extended to 500 psia with a negligible impact

~

t Therefore, use of a 1.33 limit DNBR on predictive capability.for steamline break analyses of Westinghouse press The minimum DNBR predicted for the reactors is justified.

Callaway cycle 2 steamline break transient satisfies all DNBR design criteria.

It should be noted that the.above information was previously transmitted to Dr. C. O. Thomas, Standardization and Special (Reference 5).

Projects Branch, in No'rember of 1985 l

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WIs;;NGHOUSE PROPRIETARY CLASS 3

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TABLE 2 W-3 CRT CO'RRELATION STATISTICS

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Sample Standard I

Pressure Range Number of (psia)

Data Points M/P De'riatien Limit DNER 1000 - 2300 809 0.996 0.132 1.30 500 - 2300 921 1.010 0.150 1.33 e

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WESTINGNOUSE PROPRIETARY CLASS 3 6

s EzrERENCES 1.

J. Skariska, " westinghouse Wet Annular Burnable Abso'rber Evaluation Report", WCAP-10021, February 1982.

2 2.

ULNRC-1227 dated 12/13/85.

3.

Tong, L. 'S., " Prediction of Departure from Nucleate Boiling for Axially Non-Uniform Meat Flu.M Distribution," J. Nuclear Enerev, Vol. 21, pp 241-248 (1967).

4.

Owen, D. B. " Factors for One-Sided Tolerance Limits and for variable Sampling Plans," SCR-607, March 1963.

~

5.

Letter from E. P. Rahe,.Jr., Westinghouse to Dr. C. 0.

Thomas, USNRC, " Westinghouse Response to NRC Request No. 3 on WCAP-9226-P/ WCAP-9227-N-P", NS-N'C-85-3079, November 1985.

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