ML20133J251

Affidavit of Yi-Hsiung Hsii Addressing Contention (D) Re Three Issues Stated in ASLB 850816 Order Denying Licensee Motion for Summary Disposition

ML20133J251
Person / Time
Site:	Turkey Point
Issue date:	10/15/1985
From:	Hsii Y Office of Nuclear Reactor Regulation
To:
Shared Package
ML20133J243	List: ML20133J240 ML20133J251
References
OLA, NUDOCS 8510180347
Download: ML20133J251 (10)

Text

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i l j UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l

i BEFORE THE AT0FIC SAFETY AND LICENSING BOARD

. In the Matter of

! Docket Nos. 50-250 OLA-1 t FLORIDA POWER & LIGHT COMPANY 50-251 OLA-1 l (VesselFluxReduction)

(Turkey Point Plant, Units 3 and 4) l I l

J AFFIDAVIT OF YI-HSIUNG HSII J REGARDING CONTENTION (d)

J j

l I, Yi-Hsiung Hsii, being duly sworn, state as follows:

(

i 1. I am a Nuclear Engineer in the Core Performance Branch of the  ;

t Division of Systems Integration in the Office of Nuclear Reactor  !

Regulation, U.S. Nuclear Regulatory Comission. A summary of my  ;

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professional qualifications and experience is attached to my September 4, 1

1984 affidavit which was filed as part of the NRC Staff Response to Licensee Motions for Sumary Disposition of Contentions (b) and (d),

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dated September 4, 1984  !

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, 2. The purpose of this Affidavit is to address Contention (d) with

! regard to the three issues stated by the Licensing Board in its l August 16, 1985 Order denying the Licensee's Motion for Sumary Disposi- l i

tionofContention(d). I have read the " Licensee's Motion for Sumary DispositionofIntervenors' Contention (d)"andthe" Licensee'sStatement i

of Material Facts as to Which There Is No Genuine issue to be Heard With

RespecttoIntervenors' Contention (d),"datedSeptember 20, 1985. The i

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material facts stated in relation to Contention (d) are correct and I concur in the conclusions reached in the supporting affidavit.

3. As set forth by the Licensing Board, the three genuine issues as to material facts which remain for litigation are:

(i) Whether the DNBR of 1.17 which the amendments impose on the OFA fuel in Units 3 and 4 compensates for the three uncertainties outlined by the Staff in its December 23, 1983 SER on the an,endments, at 4.

(ii) Whether, if the DNBR of 1.17 does not compensate for those uncertainties, the SRP's 95/95 standard, or a comparable one, is somehow satisfied.

(iii) Whether, if that standard is not being satisfied, the reduction in the margin of safety has been significant.

Order at 64 Before addressing the specific issues raised by the Board, the following discussion explains how departure from nucleate boiling ratio (DNBR) is detemined, the differences between the capability of the W-3 and WRB-1 correlations to predict critical heat flux and the uncertainties imposed on the Licensee's calculated minimum DNBR for normal operation and anticipated operational occurrences.

4. To support the amendment, the licensee performed a plant speciffe safety analysis to determine the minimum DNBR. As discussed in ey previous affidavit (Paragraphs 5-7), DNBR is defined as the critical heat flux divided by the actual heat flux. Criticalheatflux(CHF)is the maximum heat flux occurring just before a change of boiling heat transfer mode results in a fuel cladding temperature excursion. The CHF for the Low' Parasitic (LOPAR) fuel is calculated with the W-3 L-Grid CHF correlation. The CHF for the Optimized Fuel Assembly (OFA) fuel is calculated using the WRB-1 correlation. Both W-3 and WRB-1 correlations have been approved for safety analysis with the required DNBR limits of

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1 1.3 for the W-3 correlation and 1.17 for WRB-1 correlation. Each of 1

, these DNSR limits is imposed on the respective CHF correlation as the e specified acceptable fuel design limit to ensure with a 95 percent

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! probability at 95 percent confidence level, as specified in NUREG-0800, i

" Standard Review Plan" (SRP), Section 4.4, that the hot fuel rod in the '

core will not experience departure from nucleate boiling (DNB) during normal operation and anticipated operational occurrences.

5. Reactor operation is restricted so that actual heat flux is j

! below the CHF to ensure that the fuel does not experience DNB during j normal operation and anticipated operational occurrences. CHF is calcu- {

! r l lated using empirical correlations developed based on experimental CHF  :

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data. The W-3 correlation, an older correlation developed from CHF tests conducted with water flowing inside heated tubes, was later modified and ,

j designated as the W-3 L-Grid correlation to apply to the test results j representative of the L-Grid LOPAR fuel design. The WRB-1 CHF, a more .

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recent correlation, was developed based on the CHF test data of the rod l

bundle representative of the reactor fuel assembly geometry and operating f

ranges. ,

.! 6. DNBR limits greater than 1.0 are required to account for l variations in the data used to develop the respective correlations for CHF l predictions. These DNBR limits ensure with a 95 percent probability at

) 95 percent confidence level that DNB will be avoided when DNBR calculated

, with all uncertanties accounted for in the analysis is greater than these l va ues. The lower DNBR Ifmit of 1.17 for the WRB-1 correlation reflects

! a correlation more capable of predicting CHF than the W-3 correlation based on a better understanding and correlation formulation of the CHF phenomenon and an improved CHF test facility that result in more accurate i

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j measured CHF data. Therefore, imposing a DNBR limit of 1.17 for WRB-1 as the specified acceptable fuel design limit provides the same assurance as  ;

the DNBR limit of 1.3 for W-3 and meets the acceptance criterion of the l j SRP.  !

i 7. In the staff's December 23, 1983 SafetyEvaluation(SE),three .

uncertainties are identified that the Licensee's calculated DNBR value

! must make allowance for above the specified acceptable fuel design limit of 1.17. These are: (i) a rod bow penalty of 5.5 percent of DNBR,

! (ii) a transitional mixed core penalty of 3 percent and (iii) an uncertainty of less than 2 percent for the application of the WRB-1 l correlation to the 15x15 0FA fuel design. SE 9 3. .

l j 8. The rod bow penalty accounts for the fact that fuel rod bowing '

i results in reduction of the critical heat flux and therefore reduction in  !

DNBR. Since the Licensee's safety analysis was perfonneu u"hout the  !

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assumption of fuel rod bowing in the input to the computer program, the  ;

i resulting calculated DNBR should be reduced by 5.5 percent to account for  ;

ll  !

the rod bow uncertainty, i

9. The mixed core penalty accounts for the fact that coexistence of two different fuel designs having different hydraulic resistance l j characteristics affects the cross flow between the different fuel bundles in such a way that the fuel design having the higher grid resistance will have less flow. Since the OFA fuel hcs higher grid resistance, more flow l would be diverted to the LOPAR fuel. Since the plant specific safety  !

) analysis was performed with the assumption of either a whole core of 0FA l

l or a whole core of LOPAR fuel, a penalty is applied to the OFA analysis

! results to account for this decreased flow. In other words, the DNBR I

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calculated for a whole core of 0FA is reduced by the 3 percent mixed core peralty. No penalty is applied to the LOPAR fuel since a mixed core configuration is advantageous to LOPAR fuel in that more flow is diverted into the LOPAR fuel.

10. With regard to the application of the WRB-1 CHF correlation to the 15x15 0FA fuel, the staff required an uncertainty because of the lack of 15x15 0FA CHF data. This uncertainty was imposed on Turkey Point even though the WRB-1 correlation had been approved for the 17x17 0FA and additional CHF data was submitted by Westinghouse for the 14x14 0FA to support the application of WRB-1 to the 15x15 0FA. The Staff did not

quantify the exact value of uncertainty except to indicate that the uncertainty would be less than 2 percent. The inclusion of the 2 percent uncertainty in Turkey Point was necessary because the staff's review of the applicability of WRB-1 to 15x15 0FA was not complete at the time of Turkey Point application. However, subsequent to the issuance of the contested amendments, the staff evaluated the 14x14 0FA CHF data and concluded that WRB-1 is applicable to both 14x14 0FA and 15x15 0FA fuel designs with the same DNBR limit of 1.17. This is documented in a Safety Evaluation, dated June 29, 1984 Letter to E. P. Rahe, Jr. of Westing-house, " Acceptance of Referencing of Licensing Topical Report -

WCAP-8762(P)/WCAP-8763(NP), Supplement 1, ' Basis for Applicability of the WRB-1correlationto15x150FAand14x140FA'"(attached).

11. In response to the Board's first questian, whether the DNBR limit of 1.17 which the amendments impose on the OFA fuel in Turkey Point Units 3 and 4 compensates for the three uncertainties outlined by the

.i staff in its December 23, 1983 SER on the amendments, the answer is no.

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. I i i j The DNBR limit of 1.17 is strictly the limit derived from the measured to predicted CHF ratio data to ensure that there is a 95 percent probability at a 95 percent confidence level that DNB will not occur. This is l considered a specified acceptable fuel design limit for nonnal operation and anticipated operational occurrences as discussed in General Design l

I Criterion 10 of 10 CFR 50, Appendix A.

l The DNPR limit of '.17 does not compensate for the uncertainties 1

I associated with rod bow, the hydraulics of mixed core or the application

! of the WRB-1 correlation. These items were not among the assumptions included in the input to the predictive analysis used to calculate mini- ,

i mum DNBR. Generally, there are two approaches to account for uncertain-

ties not included in the analysis

1) the DNBR specified acceptable fuel 1

design limit may be increased by an amount equal to the penalty, or

2) the minimum DNBR ce'culated by predictive computer analysis may be

) reduced or penalized by an amount equal to the uncertainty penelty. In  ;

l either case, the calculated minimum DNBR, even if penalized for all l uncertainties as in the second method, must exceed the DNBR specified i

acceptable fuel design limit. As indicated in Section 3 of the SE on the l amendments, the second approach was used to determine the minimum DNBR i

for Turkey point. That is, the specified acceptable fuel design limit j (the 95/95 limit) of 1.17 does not compensate for the three uncertainties i stated. Rather, these uncertainties (a total of 10.5%) are compensated ,

l for by the 12.7% margin between the calculated minimum DNBR of 1.34 and the 95/95 limit of 1.17. In other words, after subtracting the 10.5%  !

penalty for the three uncertainties from the calculated DNBR of 1.34, the final DNBR (1.20) is still greater than the 95/95 DNBR limit of 1.17.

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12. In response to the Board's second question, whether, if the DNBR limit of 1.17 does not compensate for these uncertainties, the Standard Review Plan's 95/95 standard, or a comparable standard is sorrehow satisfied, the answer is yes, the 95/95 standard is satisfied.

As stated above, a DNBR limit of 1.17 with the WRB-1 correlation meets the 95/95 standard. If the calculated minimum DNBR, after accounting for

uncertainties, is greater than the DNBR design limit of 1.17, the SRP's 95/95 standard is also satisfied.

The licensee's plant specific safety analysis was performed using a

! calculated minimum DNBR greater than the 95/95 Ifmit of 1.17 (i.e., 1.34).

A DNBR of 1.34, which is referred to as the calculated minimum DNBR, was derived from the licensee's predictive analysis to provide a margin above the 1.17 95/95 limit. This margin can be used for future core altera-tions and to corrpensate for uncertainties which have not been accounted for in the input to the plant specific safety analysis. As discussed in paragraph 11, the Staff imposed three uncertainties because the effects of rod bow, the transition core and the applicability of WRB-1 correlation were not accounted for in the Licensee's input to the Turkey Point safety analysis. The 5.5 percent for the rod bow penalty, 3 percent for the transitional mixed core configuration and 2 percent for the application for the WRB-1 correlation to the 15x15 0FA fuel design result in a required total DNBR penalty of 10.5 percent for Turkey Point Units 3 and'4.

Subsequent to the issuance of the amendment, our final evaluation of the application of the WRB-1 correlation to the 15x15 0FA fuel determined I

that WRB-1 is applicable to the 15x15 0FA with a DNBR limit of 1.17.

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l Thus, the penalty previously required for the uncertainty associated with i

l the application of the WRB-1 correlation is no longer necessary and the  ;

1 l total penalty is actually lower than the 10.5 percent cited in the safety i  ;

j evaluation. Even assuming the appropriate penalty is 10.5 percent, the -

Licensee's use of 1.34 as the minimum DNBR limit provides 12.7% margin relative to the 95/95 limit of 1.17. This margin is more than enough to I

compensate for the 10.5 percent penalty and assures that the 95/95 l.

j standard is met.

] 13. The answer to the Board's third question is that DNBR of 1.17  !

j meets SRP's 95/95 standard and a 1.34 calculated minimum DNBR is more

) conservative (even if penalized for uncertainties). There has been no  ;

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j sigrificant reduction in the margin of safety and, in fact, no reduction i 1

! below the minimum safety margin provided by the 95/95 standard. ,

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j The conclusion in staff's previous affidavit that, "Therefore, the  !

j DNBR limit of 1.17 for WRB-1 as applied to the Turkey Point 15x15 0FA l l

does not result in significant reduction in safety margin," which the i Board questions in its Order (at 52-53), is based on the following: l 1

(a) The DNBR limit of 1.17 for WRB-1 is equivalent to the DNBR limit of 1.3 for W-3 in that both limits represent the 95/95 ,

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j standard specified in SRP; j i

l (b) Application of WRB-1 and the 1.17 DNBR limit to 15x15 0FA has been found acceptable without any uncertainty; and l

l (c) Even if an uncertainty of less than 2 percent is assigned for '

l the application of WRB-1 to the 15x15 0FA, as previously stated i

j in the staff SE for Turkey Point Amendment, this uncertainty I l

l together with the rod bow and mixed core penalties are j l

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compensated by the 12.7% margin between the calculated minimum 1 ,

j DNBR limit of 1.34 used by the licensee and the DNBR limit of [

l 1.17 for WRB-1.

Based on these observations, there is no reduction in safety margin from the use of the DNBR limit of 1.17 for WRB-1 in the Turkey Point l 15x15 0FA fuel design, t

l The DNBR limit of 1.17 for WRB-1 in comparison to the DNBR limit of -

i l 1.3 for W-3 has been interpreted by the Intervencrs as a significant j reduction in safety margin. However, the safety margin is not determined i 1 .

j by a specific value of DNBR limit but a DNBR limit which provides a {

l protection against DNB with a 95/95 probability / confidence standard. <

Since both 1.17 for WRB-1 and 1.3 for W-3 meet the 95/95 standard and i

! provide the same degree of assurance that DNB will not occur, the use of l a lower DNBR limit resulting from better experimental data and better CHF i correlations which yield better predictions of the occurrence of DNB >

results in no reduction in safety margin provided by the 95/95 standard.

l An adequate margin of safety is provided by a DNBR limit of 1.17 for I

WRB-1 and the 15x15 0FA fuel design at Turkey Point because the 1.17 DNBR l limit meets the SRP's 95/95 criteria. Furthermore, because the licensee's ,

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! calculated minimum DNBP of 1.34, even if penalized for uncertainties 1

l totalling 10.5%, is still greater than the 1.17 specified acceptable

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fuel design limit, the minimum DNBR limit of 1.34 more than meets the

) 95/95 standard.  !

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In summary, the margin of safety is not significantly reduced because there is no reductin in safety margin provided by the 95/95 standard. ,

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1 The foregoing is true and correct to the best of my knowledge and belief.

Yi-Hsiung Hsii Subscribed and sworn to before me this g day of October, 1985 M.dded Notary Public My comission expires: 73/t6 I <

((j

-(g [ g UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 1

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M 2 91984 Mr. E. P. Rahe, Jr., Manager Nuclear Safety Department Water Reactor Divisions Westinghouse Electric Corporation P. O. Box 355 Pittsburg, Pennsylvania 15230

SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REFORT -

WCAP-8762(P)/WCAP-8763(NP), SUPPLEMENT 1 " BASIS FOR THE APPLICABILITY OF THE WRB-1 CORRELATION TO 15 x 15 0FA AND 14 x 14 0FA FUEL" We have completed our review of the subject topical report submitted November 18, 1983, by Westinghouse Electric Corporation. We find the report to be acceptable for referencing in license applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation, which is enclosed. The evaluation defines the basis for acceptance of the report.

We do not intend to repeat our review of the matters described in the report and found a.cceptable when the report appears as a reference in license applications, except to assure that the material presented in applicable to the specific plant involved. Our acceptance applies only to the matters described in the report.

In accordance with procedures established in NUREG-0390, it is rec;ested that Westinghouse publish accepted versions of this report, proprietary and non-proprietary, within three months of receipt of this letter. The accepted versions shall incorporate this letter and the enclosed evaluation between '

the title page and the abstract. The accepted versions shall include an A (designating accepted) following the report identification symbol.

Should our criteria or regulations change such that our conclusions as to the acceptability of the report are invalidated, Westinghouse ar.d/or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of the topical report without revision of their respective documentation.

Sincerely, l

Gd4 O. f ==:

Cecil 0. Thomas, Chief I

y , ,,,a Standardization and Special .

9-m,,w % Projects Branch Division of Licensing

Enclosure:

As stated . . . _ _ _ _ _ . . - - . . - - . . . . _ _ .

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ENCLOSURE WCAP-8762(P)/WCAP-8763(NP)

SER ON THE APPLICABILITY OF WRB-1 TO WESTINGHOUSE 14X14 AND 15X15

1. INTRODUCTION AND BACKGROUND The Westinghouse rod bundle critical heat flux (CHF) correlation, WRB-1, has prevously been approved (Ref.1,2) for application to the 15x15 and 17x17 standard low parasitic (LOPAR) fuel assemblies and 17x17 optimized fuel assemblies (OFA). The approved minimum departure fran nucleate boiling ratios (DNBR) are 1.17 for the standard R-type mixing vane grid assembly and the 17x17 0FA, and 1.37 for the standard L-Grid assembly.

By letter dated November 18,1983 (Ref. 3), Westinghouse submitted Supplument 1 to WCAP-8762, " Basis for the Applicability of the WRB-1 correlation to 15x15 and 14x14 0FA Fuel " to apply for the extension of the application of WRB-1 to the 14x14 and 15x15'0FA with a DNBR limit of 1.17.

The differences between the OFA and standard LOPAR fuel are the fuel pin diameters and the mixing vane spacer grid designs. The 17x17 0FA has a pin outer diameter of 0.36 inches versus 0.374 inches for the standard 17x17 fuel; the 14x14 0FA and 15x15 0FA have pin diameters of 0.40 inches and 0.422 inches, respectively, compared to 0.422 inches for both 14x14 and 15x15 standard fuel. The OFA spacer grid has thicker and higher straps made of Zircaloy having different mechanical properties than the Inconel straps used in the standard grids. Although the OFA grid dimensions are different than the standard LOPAR grid, Westinghouse stated that there is no significant departure from the type-R grid characteristics.

The WRB-1 correlation as described in WCAP-8762 (Ref. 4) was orginally approved for application to the 17x17 and 15x15 standard, fuel assenblies l based on the available CHF test data representative of standard fuel.

t 6.

In order to extend the application of WRB-1 to the 17x17 0FA, Westinghouse in its submittal WCAP-9401 (Ref. 5) provided additional CHF test data fran a test assably representative of the 17x17 0FA. To justify that the DNBR limit of 1.17 is applicable to the 17x17 0FA, a statistical analysis  ;

as perfomed to show that the OFA data belong to the same population as the standard R-grid data. The result of this analysis showed that the OFA data were within the tolerance limits of the means of the measured to predicted CHF ratios (M/P) of the comparable standard R-gria data. The '

analysis of variances showed that the null hypothesis, that the variances i

of the M/P ratios of the OFA typical cell and standard R-grid typical cell are equal, would not be rejected at a 5 percent significance level. For the OFA thimble cell data, a similar hypothesis would be rejected at a 5 percent significance level. However, since the rejected 0FA data had a lower standard deviation, it as shown that a DNBR limit of 1.17 would be conservative when applied to the rejected data. Based on this analysis, WRS-1 was found acceptable for application to the 17x17 0FA with the same DNBR 1imit of 1.17.

The extension of the WRB-1 applicability to the 14x14 and 15x15 0FA will be evaluated and discussed in the following section.

2. STAFF EVALUATION 1

) i l In order to justify the extension of the applicability of WRB-1 to the i 14x14 and 15x15 0FA, Westinghouse provided additional CHF test data from l the test bundle representative of the 14x14 0FA. Westinghouse maintains that the same DNBR limit of 1.17 for the standard R-grid fuel is also applicable to the 14x14 and 15x15 0FAs. To justify this assertion, it i

must be shown that the 14x14 0FA data belong to the same population  ;

which as used to determine the limit, or that 1.17 was a conservative  :

limit relative to any limit based on the 0FA data. In addition, since ,

l there is no CHF data representative of the 15x15 0FA, it- is necessary  ;

that the 15x15 0FA geanetry is within the WRB-1 applicability range. l l

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'. - l In order to show that the 14x14 0FA test data belong to the same population as the standard fuel data, Westinghouse has perfomed a statistical analysis of the measured to predicted CHF ratios fra the 14x14 0FA test data and the caparable standard R-grid data with fuel OD of 0.422 inches. The analysis consists of a F-test for equality of population variances and a t-test for equality of population means.

The results show that both null hypotheses of equal variances and means of the 14x14 0FA and the comparable standard R-grid data distributions cannot be reaected at a 5 percent significance level. In this analysis.

Westinghouse has combined the 2 sets of standard R-grid data having 0.422 inches pin diameter and 26 inches grid spacing into one group and therefore the tests of variances and means are simply two group analyses between the standard and 0FA data. The staff has perfomed a one way analysis of variances for the equality of means test as an independent check. In this calculation, the standard R-grid data are not combined into one group. The result of this staff analysis of variance also shows that the null hypothesis of equal means cannot be rejected at 5 percent significance level . Therefore, the 14x14 0FA CHF data can be incorporated into the total R-grid population. Using the mean and standard deviation of the (M/P) ratios corresponding to the combined total data, the derived one-sided tolerance minimum DNBR limit would be 1.163 with 95 percent probability at a 95 percent confidence level of avoiding DNB. In addition, the 95/95 DNBR limit derived from the 14x14 0FA data alone would be 1.122.

These DNBR limits are below the proposed 1.17. Therefore, the staff concludes that WRB-1 is applicable to the 14x14 0FA with a DNBR limit of 1.17.

The 15x15 0FA design is virtually identical to the 15x15 R-grid design except for the spacer gria dimensions. In order to minimize the effect of the grid dimensional changes on DNB perfomance, Westinghouse stated that special care was taken in the OFA grid design to preserve the important mixing vane characteristics, flow area ratio, and the dimple and spring arrangement used for maintaining the grid-to-rod contact. The

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success of the use of this scaling technique in the OFA grid design has been ' demonstrated by the fact that WRB-1 predicts both 17x17 and 14x14 0FA CHF data well without any modification to the correlation. Repeata bility studies performed as described in WCAP-9401 have shown that the WRB-1 correlation's prediction capability is essentially identical for both 0FA and standard fuel geanetry, indicating no additional conponent of variance is introduced by the grid dimensional change. In other words, the correlation correctly accounts for the equivalent diameter effects and the scaling approach for the grid dimensional changes.

Since the 15x15 0FA diameter, rod pitch, heated length and grid spacing are within the WRB-1 applicaDility range and the success of the scaling technique in the 0FA mixing vane grid design has been proved by the similar 14x14 and 17x17 0FA grid design, we conclude that WRB-1 is also applicable to the 15x15 0FA with a DNBR limit of 1.17.

3. REGULATORY POSITION The staff has reviewed Supplement I to WCAP-8762. We find that the WRB-1 correlation is acceptable for application to both 14x14 0FA and 15x15 0FA with a minimum DNBR limit of 1.17. This acceptability is subject to other restrictions which were imposed in the staff safety evaluation reports (Refs. I and 2) on WCAP-8762 and WCAP-9401.

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REFERENCES

1. Memorandum from D. F. Ross, Jr. (NRC) to D. B. Vassallo (NRC), " Topical Report Evaluation for WCAP-8762," April 10,1978.

2. Letter from R. L. Tedesco (NRC) to T. M. Anderson (Westinghouse),

" Acceptance for Referencing Topical Report WCAP-9401(P)/WCAP-9402 (NP),"

May 7, 1981.

3. Letter from E. P. Rahe, Jr. (Westinghouse) to C. O. Thomas (NRC), " Basis for Applicability of the WRB-1 CHF Correlation to 15x15 0FA and 14x14 0FA Fuel, Supplement 1, WCAP-8762," NS-EPR-2854, November 18, 1983.

4. F. E. Motley, et al, "New Westinghouse Correlation WRB-1 For Predicting Critical Heat Flux In Rod Bundles With Mixing Vane Grids," Westinghouse Topical Report WCAP-8762, July 1976.

5. M. D. Beaumont and J. Skaritka, " Verification Testing and Analyses of the 17x17 Optimized Fuel Assenbly," WCAP-9401, March 1979.

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

Docket Nos. 50-250 OLA-1 FLORIDA POWER AND LIGHT COMPANY l 50-251 OLA-1

)

(Turkey Point Plant, Units 3 and 4)) (Vessel Flux Reduction)

CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE TO LICENSEE'S SECOND MOTION FOR

SUMMARY

DISPOSITION OF CONTENTION (d)" in tile above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or as indicated by an asterisk, by deposit in the Nuclear Regulatory Commission's internal mail system, this 15th day of October, 1985:

• Dr. Robert M. Lazo, Chairman Noman A. Coll, Esq.

Administrative Judge Steel, Hector & Davis Atomic Safety and Licensing Board 4000 Southeast Financial Center U.S. Nuclear Regulatory Comission Miami, FL 33131-2398 Washington, DC 20555

• Atomic Safety and Licensing Board

• Dr. Emeth A. Luebke U.S. Nuclear Regulatory Commission Administrative Judge Washington, DC 20555 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Comission

• Atomic Safety and Licensing Washington, DC 20555 Appeal Board (8)

U.S. Nuclear Regulatory Comission

• Dr. Richard F. Cole Washington, DC 20555 Administrative Judge Atomic Safety and Licensing Board

• Docketing & Service Section U.S. Nuclear Regulatory Commission Office of the Secretary Washington, DC 20555 U.S. Nuclear Regulatory Comission Washington, DC 20555 Harold F. Reis, Esq.

Newman & Holtzinger, P.C. Joette Lorion 1615 L St;, NW 7269 SW 54th Avenue Washington, DC 20036 Miami, FL 33143 Martin H. Hodder, Esq.

113I N. E. 86th Street Miami, FL 33138 Y* YW Mitzi't ' Young " V Counsel for NRC Staff