Text

Qualification of WRB-1 CHF Correlation in VEPCO Cobra Code
	ML20059L076
Person / Time
Site:	Surry, North Anna
Issue date:	07/31/1990
From:	Basehore K; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML18153C367	List: ML18153C366; ML20059L076;
References
	VEP-NE-3-A, NUDOCS 9009260098
	Download: ML20059L076 (69)
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VP-NE-3-A B

QUALIFICATION OF THE WRB-1 CHF CORRELATION IN THE VIRGINIA POWER-COBRA CODE I

by I

R. C. Anderson N.

P. Wolfhope I

Nuclear Analysis and Fuel

' Nuclear Engineering Services-y Virginia Power I,

July, 1990 I

Recommended for Approval:

,I K.'L.'Basehore, Supervisor Nuclear Safety Analysis

^

Approved by:

=

R. M.'Berryman, Manager Nuclear-Analysis and Fuel I

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g UNITED STATBS NUCLEAR REGULATORY C004All8810N I5 WAENNeeTeN o,c.sEEEs

(

July 25, 1989 l

Docket Nos. 50-280, 50-281, 50-338 and 50-339

'I Mr. W. R. Cartwright NMAE Vice President - Nuclear I

Virginia Electric and Power Company Nuclear Operations 5000 Dominion 01vd.

g ec 7].

Glen Allen, Virginia 23060

,, Q,--

g,-

Dear Mr. Cartwright:

I

SUBJECT:

SURRY UNITS 1 AND 2, ANO NORTH ANNA UNITS 1 AND 2 - USE OF VIRGINIA POWER TOPICAL REPORT VEP-NE-3, " QUALIFICATION OF THE WRS-1 CNF -

CORRELATION IN THE VIRGINIA POWER C08AA CODE" (TAC NOS. 67363, 67364, 71071 AND 71072)

By letter dated January 29, 1987, you requested apsreval of the above.

subject Topical Report VEP-NE-3 to use the WRS-b correlation to perfore Departure from Nucleate Soiling Ratio analyses for the Surry and North Anna Power Stations.

Based on our review, with assistance from our consultant, Pacific Northwest L

Laboratories, we conclude that the VEP-NE-3 Topical Report is acceptable for-application at the Surry and North Anna Power Stations with the constraints.

that the. critical heat fluxes shall not exceed 1.0 letu/hr-ft and that the einieue grid spacing shall be 13 inches er greater.

Stacerely, g

gg Sus C. Laines, Assistant Director

E-for Region II Reactors 5:

Division of Reactor Projects I/II.

Office of Nuclear Reactor Regulation

Enclosure:

Safety Evaluation f l '-

cc w/ enclosure:

Ea See next page LI

~

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...=

Mr. W. R. Cartwright I

North Anna Power Station, Units 1&2 Surry Power Station, Units 1&2--

cc:

Mr. William C. Porter, Jr.

C. M. G. Buttery, M.D., M.P.H.

County Administrator Department of Health Louisa County 109 Governor Street P.O. Box 160 Richmond, Virginia 23219 Louisa, Virgin n 23093 4

Re Michasi W. Maupin, Esq.

U.gional Administrator, Region II S. Nuclear Regulatory Commission Hunton and Williams 101 Marietta Street N.W., Suite 2900 I

~Richmoud, Virginia 23212 P. 0. Box 1535 Atlanta, Georgia 30323 Mr. W. T. Lough

i Virginia Corporation Commission Mr. G. E. Kane Division of Energy Regulation P. D. Box 402 P. O. Box 1197 Mineral, Virginia 23117 I

Richmond, Virginia 23209 Mr. W. R. Cartwright Old Dominion Electric Cooperative Vice President - Nuclear c/o Executive Vice President Virginia Electric and Power Comparty I

Innsbrook Corporate Center 5000 Dominion 81vd.

.~.

4222 Cox Road, Suite 102' Glen Allen, Virginia 23060-Glen Allen, Virginia 23060 I=

Mr.- Michael Kansler, Manager Mr. W. L. Stewart Surry Power Station Sinfor~Vice President - Power Post Office Sox 315 1

l Virginia Electric and Power Co.

Surry, Virginia 23883 Post Office Box 26666 Richmond, Virginia 23261 Resident Inspector Surry Power Station

- l Mr. Patrick A. O' Mare U.S. Nuclear Regulatory Commission Office of the Attorney General Post Office Sox 166, Route 1 Supreme Court Building Surry, Virginia 23883

-I-101 North 8th Street Richmond, Virginia 23219 Mr. Sherlock Holmes, Chairman Board of Supervisors of I

Resident Inspector / North Anna Surry County c/o U.S..MAC.

Surry County Courthouse Senior Resident Inspector.

Surry, Virginia 23683 Route 2, Bos.78

(.

Mineral, Virytaia 23117 i

Attorney General Supreme Court Su11 ding 101 North 8th Street Richmond, Virginia 23219 I

g u

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usNTEDstATas

[:

NUCLEAR RROULATORY COMM188HNd s

m.o.a. asses I

\\.....

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l

RELATING TO TOPICAL REPORT VEP-NE-3. "0UALIFICATION OF THE WRB-1 CHF CORRELATION IN THE VIRGIN!A POWER CO M CODE" VIRGIN!A ELECTRIC AND POWER CO*ANY SURRY AND NORTH ANNA POWER STATIONS UNITS 1 AND 2 DOCKET N05. 50-280. 50-281.50-33t AND 50-339

. a'n

1.0 INTRODUCTION

'B

.By letter dated January 29, 1987, Virginia Electric and Power Company (VEPC0) submitted the Virginia Power Topical Report VEP-NE-3, " Qualification of the I

WRR-1 CHF Correlation in the Virginia Power COBRA Code" (Ref.1), for NRC review.

The purpose'of this subitta) was to qualify the WRS-1 correlation for Departure d

s from Nucleate Soiling AM,io (DER) nnalysis in order to replace the W 3 correlt-I, tion which has previously been used in DE R analyses.

The improved ac.euracy 67.

i i.no kRS-1 correlation results in a sestantial gain in Om margin over the use of the W-3 correlation.

'h The review considered only the issue of whether or not it is appropriata to substitute the VEPC0 version of the COBRA code for the Westinghouse THINC code in the theres)-hyoraulic analyses using the WRS-1 Critical Heat Flux (CHF)

.E

~ correlation. L The question of the appropriateness of the method used to -

5 detamine the Minimum Departure free Nucleata Soiling Ratio'(WER) limit for the CHF correlation, regardless of which thermel-hydraulic code is.used, was not I

addressed.-

The WRS-1 correlation and its 1988 limit were previously approved l

by the NRC.(Ref. 2),

a 2.0 EVALUATION I

l The staff has coordinated this subject review through our consultant, Pacific Northwest Lahoretaries (PNL). The Technical Evaluation Report (Ref. 3) free I

PNL provided the staff with recommendations on the subject review. Our finding indicates theb-the CIERA code is an acceptable substitute for the THINC code in 3

the theres)-fqydraulic analysis of CHF data and in WWR calculations for I.

reactor operating conditions and operational transients.

The codes solve the same. set of. conservation equations, have eeny of the same constitutive models, i

- and use similar numerical solution algorithms.- In addition, the VEPC0 thermal-hydraulic analysis mothed, as documented in VEP-FRD-33-A, (provided in response I

I to our request for additional infomation (Ref. 4)) shoue that the code is being applied in a manner consistant with its assisiptions and capabilities.

i i

I

.g.

'E VEPC0 has shown that the C0sRA code produces essentially the same values of sB measured-to-predicted (M/P) CHF-ratio for the WS-1 correlation's data base as J

obtained in calculations with the THINC code.

This was demonstrated by applying i-

.the WRB-1/ COBRA combination to the entire WRS-1 data base and comparing the predicted DNS ratios with those obtained in the qualification of WRS-1 with the L

THINC code.

Twenty-five points were deleted free ~the #+* base because of l

suspected typographical errors in the transcription i..

__.s data as repo-+:sd in EPRI-NP-2609.

Response to our request for additional _inforsation (Ref. 5) on 1

the identification of these errors indicates that the esthod used was reasonable, l

and unlikely to roult in the deletion of valid data.

Using the COBRA code will not materially change the W WR results, in comparison with those obtained with THINC. over the intended range of application of the correlation.

The report VEP-NE-3 states only that the correlation will be o

applied over tho'same range as in the THINC code.

The intended r'ange of application of the WRS-1 correlation with COBRA-!!!C/MIT' is as follows:

1440 i pressure (psia) < 2490 i

I 0.9 < mass flux (N1ba/hr-fi8) < 3.7

~

local quality < 0.30 I

local heat flux (Mtu7hr-ft8) < 1.0 l

nixing vano grid spacing > 13.0 Tnches L

Because of questions raised concerning the nonconservative behavior of the correlation of high critical heat fluxes, (i.e., above 1.0 Mtu/hr-fta), and 3-for 13" grid spacing, additional constraints were placed on the range of 5

applicability of.the WRe-1/COsRA combination. The maxim e heat flux expected for the North Anna and Surry plants is 0.82 Mtu/hr-fta, so the correlation will not be applied in the nonconservative-region when used with C08AA.

1 8

Similarly, the behavier for 13" grid spacing < s not a concern since the fuel for these plants has a grid spacing of 20" and 26", respectively. Within these constraints the COBRA code can be used with the WRB-1 CW correlation for I

analysis of the Westinghouse 17x17 standard "R" grid fuel,17x17 Vantage 5H zircaloy fuel at the North Anna plant, and the 15x15 0FA-type fuel at the-Surry plant.

3.0 CONCLUSION

VEpCO has adegastely demonstrated that the W M-1 C W correlation yields essentially the samer MMA results.when the COBRA code is substituted for the THINC code. in the thereal-hydraulic calculattens.

Therefore,.the VEP-M topical report is acceptable for VEPC0 plant-specific application with l

constraints that the critical heat-fluxes shall not exceed 1.0 Mtu/hr-fta and that no grid spacing shall be less than 13".

Dated: July 25, 1989 Prircical Contributor:

T. huang I

e

}

4.0 REFERENCES

1.

Letter, W. L. Stewart (VEPC0) to USNRC Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2. " Qualification of the WRS-1 CHF Correlation in the Virginia Power COBRA Code," VEP-NE-3, January 29, 1987.

2.

F. E. Motley, et al., "New Westinghouse Correlation WRS-1-for Predicting Crit.ical Heat Flux in Rod Sundles with Mixing Vane Grids, WCAP-8762-P-A (Proprietary) and WCAP-8763-A (Non-Proprietary), July 1984.

3.

J. M. Cuta'and W.. K. Winegardner, " Review of Virginia Electric and Power Corporation submittal VEP-NC-3, " Qualification of the WRB-1 CHF I

Correlation in the Virginia Power COSRA Code," March 1989.

Memorendum,'M. W. Hodges to H.' N. Berkow,- Request for Additional 4.

Information for the Virginia Power Topical Report VEP-NE-3, i

" Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," December 8, 1988.

5.

Letter, W. R. Cartwright (VEFCO) to USNRC, Virginia Electric and Power-i Company North Anna Power Station Units 1 and 2, Surry Power Station Unita, I and 2, Topical Report VEP-NE-3, " Qualification of the WRS-1 CHF

' Correlation in the Virginia. Power efesA rade," December 19, 1988.

1 y3 ll la L

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1 PAGE 2

I CLASSIFICATIOM/ DISCLAIMER The

data, information, analytical techniques and conclusions in' this. report have been prepared solely for use by Virginia power (the company),

and they may not be appropriate for use in situations other than those for which they. were specifically prepared.

The Company therefore makes no claim or warranty whatsoever, express or implied, as to their accuracy, usefulness, 1

or applicability.

In particular, THE COMPANY MAME5 M0 WARRANTY OF-MERCHANTABILITY OR FITNESS FOR A PARTICULAR' PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USAGE OF TRADE, with respect to this report or any of the data, information, analytical techniques, or conclusions in-it.

By making this report I

-available, the Company does not authorise its use by others, and any such use is expressly forbidden except with the prior uritten approval of the company.

Any such written approval shall itself be deemed to incorporate the disclaimers of' liability and disclaimers

- of warranties provided herein.

In no-event shall the Company be

-liable, under any legal theory whatsoever (whether contract, tort,

warranty, en strict or absolute liability),

for any property

damage, mental or physical in:lury or

death, loss of-use of g

property, em other damage resulting from or arising out of the use, authorised or unauthorised, of this report or the

data,

,information, and analytical techniques, or conclusions in it.

I I

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El PAGE-3 i

ABSTRACT The database which was used to qualify the WRB-1 CHF correlation 'in the Westinghouse THIMC code for 17x17 and 15x15'"R" grid fuel has been re-analysed by WRB-1 with the Virginia Power COBRA code.

Data analysis has shown that DMB-protection is-provided by a 1.17' COBRA /WRB-1 DNBR limit with 952 probability at L 4

952 confidence -level.

These-results have also been shown to be applicabl'e to 14x14, 15x15 and 17x17 Westinghouse OFA-type' fuel.

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PAGE 4

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ACMM0WLEDGEMENTS I

1

-The authors would like to thank Mr.

M.

L.-Basehore'for his-

considerable technical assistance in this project.

Special thanks also go to Mrs.

A.

5.

Pegram, Mr. K.

L. Maddock and Mrs.

J.

Y.

Reynolds. for their work in the entry and verification e' the data.

I The assistance'of Dr.

3.

G.-Reddy of Columbia University and Dr.

A.

J.

Friedland of Westinghouse in-identifying and correcting data errors-was also greatly appreciated.

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.PAGE 5.

LI TABLE OF CONTENTS v

I Page Title-Page.....-...............................................

1-classification / Disclaimer....................................

't Abstract.....................................................

3 Acknowledgements.............-................................--

4 Table of Contents............................................

5 I

List of F1gures..............................................

6

~ List of Tables.................'.............................s 7

Nomenclature.................................................

8 1.0 Introduction.............................................

9 2'.0 Description of Correlation...............................

10 3.0 Data Analysis.............................................

12 I

4.0 Results..................................................

14' 5.O Application of WRB-1 Correlation.........................

26 I

6.0 conclusions...................................:...........

24 l=

References...................................................

29 Appendia &r Data Listing......................................

30 I

t 1

I 9

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PAGE, 6

I I

I LIST OF FIGURES Page 1

4.1 Predicted vs. Measured Critical Heat Flux (million BTU /hr-fts)...................................

16 l

I 4.2 Measuredato-Predicted Critical Heat Flux Ratio l

vs. system Outlet Pressure (Psia)......................

17 4'.3 Measured-to-Predicted Critical Heat' Flux Ratio I

vs.-Core-Average Mass Flux (alllion Iba/hr-fta),,,,,,,,

tg 4.4 Measured-to-Predicted Critical Heat Fluu Ratio

-l vs. Hot Channel outlet 9uality.........................

19 4.5 Measured-to-Predicted Critical Heat Flux Ratio vs. Inlet TenPerature

(*F)..............................

to 4.6 Histogram of Measured-to-Predicted CHF Ratios..........

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1 LIST OF TABLES-Jr Page 4.1 Measured-to-Predicted Critical Heat Fluu Ratio Statistics by Axial Power Profile......................

22-iI 4.2

' Measured-to-Predicted Critical Heat Fluu Ratio Statistics by subchannel cell Type.....................

22 q

+g 4.3 Me'asured-to-Predicted. Critical Heat Fluu Ratio

':g' Statistics by Rod Diameter (Inch)......................

22 j

4.4 Measured-to-Predicted Critical Heat Fluu Ratio Statistics:by Test section Heated Length.(ft)..........

23 1

4.5 Fleasured-to-Predicted critical Heat Fluu Ratio-Statistics by Grid.5 Pacing (Inches)....................

23 4.6 Comparison of summary Statistics from Virginia ll Powaz... Westinghouse and MRC' Calculations...............

24 4.7 Measu 3d-to-Predicted Critical Heat Fluu Ratio Statistics for All. Test Sections......................-

25 I

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i oAGE 4:

MOMEMCLATURE Variable ~

Definition A

Correlation coefficient A1 Correlation coefficient B

Correlation coefficient c

-33 Correlation coefficient B4 Correlation coefficient CHF Critical Heat Flu, (sillion Stu/hr-ft1)

D' Moraality test statistic Dg Distanoa to nearest upstrema grid (inch)

DNBA Departure from Mucleate Boiling Ratio I'

F Mon-uniform axial flux factor G(loc)

Local mass flux (million Iba/hr-fta) f Gsp Mixing vane grid spacing (inch) k Small sample correction factor Lh Distance from beginning of heated length to local I

node (inches) 4 M/P Measured-to-Predicted CHF ratio g

W M/F Mean Measured-to-Predicted CHF ratio OFA Optimised Fuel Assembly P

Pressure (psia) q" Heat flux (million Etu/hr-ft8) q"(CNF)

Critical Heat Flux (million Stu/hr-ft8) s' Estimated standard deviation x(loc)

Local thermodynamin quality

, I.

.I

Chaptor 18 INTRODUCTICM PAGE 9

i 7,1 1.0 Introduction f

Virginit power has recently acquired the access ti3.sts to the Westinghouse WRS-1 Critical Heat 71ux (CHF) correlation.i k CHF correlation is used in the caiculation of the Departure from I

!!ucle a te toiling Ratio (DNBR),

which is one of the key analysis criteria in many of the UFSAR Chapter 15 safety analyses.

The DNBR is the ratio of the Critical Heat Flux (CHF er g*(CHT)) and the local heat flux q".

To be licensed for use, e C87 worrelatAon av,t

' be tested against experimental data whish span the anticA' gated range of conditions over which the correlation will as applied.

Further, the population statistics of the data base aust be uses to set a DNBR limit such that the probability of avoiding DNS will be at least 952 at a 952 confidence level.

This limit is referred to-as the 95/95 DNBR limit.

The VAB-1 correlation has been qualified with the the r m al-hl'd r aulto s code COBRA: by Virginia power.

A subset of the data from the Columbia-EpRI CHF data base s fer Westi,ttyhouse "R"

grid 17x17 and 15a15 fuel was used in, %e taalyass.

Virginia potur's North Anna and surry plants utilise Westiy house l'u17 and 15x15 fuel, respectively.

This Topical Report summarises the data evaluations which were performed to qualify the codekoorrelation pair.

I 5

.4...

..,.m

.., ~... ~ _..... _ _

Chopter 28 DESCRIPTION OF CORRELATION PAGE 10 5

2.0 Description of Correlation i

AefereAse 1 begins its discussion of the WR8-1 norrelation by noting that Westinghouse mixing vane CHT data typically obey the j

functional fora q'(CHF) = A - Bsx(log)

(2.1)

' n which x(loc) is the local thermodynamic guality, and A and B are i

i functions of geometry end local conditions as follows:

A = Alp,0(loc),Lh,Gsp,Dfl

( 2. 2 )-

i

,I

~

5 = Blp.6(loo),Lhl (2.3)

Here, y

is the systen pressure, G(loo) is the local mass flux, Lh is the test section heated length. Sep is the mining vane grid i

spacing and Dy is the distance from the local node to the nearest

{

upstresa grid.

The latter two terms account for the CHF-inhibiting turbulence which the aiming vane grids induce, but which decays with distance.

Eventually, the following functional form was

{

obtained g"(CNF),a ph + A1 + 338G(loo)

I ggs:(1,,)se(los)

(2.4)

The performanea-Factor pF and the coefficients A1, 33 and 34 are functions of the independent variables.

These coefficients are-Westinghouse proprietary information and are given in Reference 1.

In

addition, the critical Heat Flum is divided by the non-uniform flum factor F,4 when a non-uniform axial power profile is used, as

,l I

I Chopter it DESCRIPT10N OF CORRE1ATION PAgt 11 I

a part of the calculation of the DNBR t*(CHF)

DNBR * ---------

(2.5) t'

T E

i in which g*(CHF) is given by equation (1.4), the local heat flux g*

j is the product of the core-average heat fivu and the local radial and axial power factors, and F is given in Reference 4.

j I

The WRB-1 cot elation uncertainty is relatively small, as

]

[

reflected in its low 95/95 DNBR limit (which is only 1.17 for l

THIMC/WRB-1).

This improved accuracy results in a substantial gain 1

in DNS margin over the use of the W-3 correlation, which has previously been employed by Vi::sinia power.

.I The COBRA /WRB-1 qualification data analysis and results are q

I l

presented on the following pages.

]

I J

B I

l I

I I

I 9

Chapter 3:

DATA AMA1YSIS PA8E 12 i

'I 3.0 Data Analysis Extensive critical Heat Flux caleviations have been performed to qualify WRB-1 in COBRA.

The modeling philosophy of Reference 2 l

I l

was employe6 throughout the project.

In these analyses. COBRA models of the appropriate EPRI test sections were built and used to analyse the Re arth 3 data.

l i

L The subst,a al thermal-hydraulies code COBRA IIIc/MIT solved l

the conservation equations to obtain local fluid conditions, which were then used to predict CHF throughout the test bundles.

These predictions were divided by the local heat fluu to obtain a DMS lI ratio.

The sussary DMBR's were stored on disk to ptsrait i

l post-processor analysis.

Since the input heat fluxes were sensured

(

Critical Heat

Flunes, the DMSR is a

reciprocal sensured-to-predicted (M/p)

CHF ratio.

Appropriate summary i

statistics were thus ganarated both for the DNBR and-the M/p ratios.

By historical precedent, the M/p statistics are used to set the correlation's 95/95 DNBR limit.

The suasary statistics L

will also be used in Virginia power's statistical DMBR Evaluation Methodology.5 I

Altheush each thermal-hydraulies code solves the same constavaties aguations, the code solution schemes and flow phenomena models any differ slightly.

It is thus necessary to

' qualify a

correlation and code as a pair, since different codes which employ the same correlation may yield different results.

l Even though COBRA /WRB-1 and THIMC.th1-1 will be shown in Section 1l l

I e

- _ _ _. -.,. ~ _ _ _. -. _ _. _ _ _. -. _. _

~_,.

Chaptor 3:

DATA ANALYSIS

-PAGE 13 4.0 to have the same 95/95 DMBR limit, it is guite possible that the same correlation could have required different DMBR limits when used in different codes.

The data from EPRI 4x4 and 5x5 "R" grid test sections were i

analysed.

The 5x5 test bundles have the same subehannel geometry as the current Westinghouse 17x17 "R" grid fuel, while the 4x4 test j

bundles have essentially a 15x15 geometry matrix.

The details of the test bundle geometries are provided in Reference 3.

]

Some of the Reference 3 data were discarded prior to their incorporation in the C03RA/WR3-1 data base.

Two criteria were used to justify data deletions.

The first criterion was consistency I

with the practice of the test sponsors certain points were f

excluded from the c03RA/WRS-1 data base because thew had been excluded from the THINC/WRB-1 data base in Reference 1.

Most excluded data were deleted under this criterion.

l The second exclusion criterion was consistency of the input f

l data in References 1

and 3

although some differences were expected.' data which differed by more than ten standard deviations L

were exclude 4-as being probable typographical errors in Reference i

(

3.

With the esseption of the 33 data points which were thrown out under the second criterion, this data base is the same as the one which was used in Reference 1 to qualify THINC/WRS-1 for "R" grid i

fuel.

I:I i

g

_..-.--~.

Chaptor 48 RRSULTS PAOR 14 l

4.0 Results g

i The DNBR data for a total of 19 test sections were collected and analysed.

consisting of 945 data points.

A plot of predicted i

versus measured Critical Heat Flux is shown in rigure 4.1, illustrating the excellent performance of the C03RA/WRB-1 combination.

The data were also examined for trends in test section geometry and operating conditions.

Plots of the M/p ratios versus

pressure, flow, thermodynamio quality and anlet temperature are I

presented in Figures 4.2-4.5.

As may be seen, no significanto 1

trends in the M/p ratic versus any of these parameters exists.

I The data are broken into subsets in Tables 4.1-4.5 to check l

I.

for trends as a

function of the axial power profile, subchannel 1

cell type, heated rod diameter, test section heated length and grid spacing.

The means and standard deviations of each subset are within a few percent of the average for the entire dataset.

We may note that the same Critical Heat Flux data were employed by Westinghouse in theirquali'f[cationofWRB-1inthe THINC code.

Fu:there the NRC performed audit calculations of some of these same data with the COBRA III and COBRR IV codes.'

The I

Westinghouse and MRC calculational results are compared to the Virginia power results in Table 4.6.

Given these results DMB is avoided with 952 probability at a 95R confidence level for "R" grid fuel by meeting the following

~

I I

l Chapter 48 RESULTS PAGE 15

\\

DNBR limit 8 I

ox..

< 4. o (M/p) - 1.645*ks s in which M/p is the mean measured-to-predicted CHF ratio, s is the estimated standard deviation of the data, the 1.645 multiplier is j

the a-value for one-sided 952 probability of a normal distribution and k

is a

small sample correction factor which provides a 95X upper confidence limit on the standard deviation.

Equation (4.1) 3 yields a

95/95 DNBR limit of. 17 for COBRA /WRB-1 Which is the I

same as the THINC/WRB-1 DNBR limit.

l Because the 95/95 DMSR limit is established by using the mathematical characteristics of the 74ernal distribution, it was desirable to verify the normality of the data.

To this and the D' normality test' was applied to the M,'y population.

roz 945 data

points, the assumption of normality cannot be rejected at a 5X level of significance if a population's calculated D' statistic lies in the range 8134.05D'58245.4.

For the C03RA/WRB-1 M/p population, the calculated D' statistic was $180s the assumption of normality is thus acceptable.

A histogram of the M/p data is b

presented in Figure 4.6.

The data statistics for each Test section are presented in i

Table 4.7.

I I

I

. - - ~.

. -.. =. -... -

. ~...... _.. -. - _. -. -. -. - -. -.

Chopter 48-RESUlfs PAGE 16 l'

16 j

)

1. 4 -

l o# e I

4-o 1 2-

g i

It i

i 1.0-l 3

+ '.

T 0.8 8

I',,

I

*g j

C n

e a.

g-H g

0.6-g i

l T

L U

e X

0.4-

,I 0.t-e

[

l 0.0 '

I 00 02 0.4 0.8 0.8 10 17 1.4 16

' =-

MEA 60 RED CRiflCAL HEAT FLUX t

I t

Figure 4.1 - Predicted vs. Measured Critical Neat Flus (million BTU /hr-fts) i I.

I I

i Cho, tor 4 RESULTS

.ASE 17 i

I i

i 1 30' 4

1 1 25-1 20-E. i.i.

I 4

5 R

g 1 10-I t

O I.05-

,I

.P R

C 1.00 0

'E C

3

{

0.95-O I

C 0 90-p t

v.

.I i

(

8 t

0..

i e*

O

0. 0 I

s 0.15-O.10 1300 1400 1600 1900 2000 2t00 2400 2600 I

PRtl6URE (P81Al

..I Figure 4.8 - Mea.ured to Fredicted Critical Neat Fium-R. tie

.............t r....r.

t 3

q l

_._-_..__.__..______.____..._____.___..________..__.._..__...__._.__3

\\

i Ch.pto.

8 RESULTS P,6E 14

I I

i i

II 4

I.

1 30 k

I 1 25-i t

i

[

j l.20-n c

.g..

e.

A l.15-oo e

[

j i

g,

d 1

E l.00 il

!0.,,.

F l.

5 r

8 i

,C 0.90-s 3

g 0.85-8,'

g i

e i

e

(

0 0.60-

)

i.

i 0.is-S e

0.10 I

05 10 15 t.0 3.5 3.0 3.5 4.0 i=

MASS FLUX (MLSM/NR/FTSSI LI

,i....... -.................... m. 1.... n.....i.

,1..

<. m i.

1,.

!I I

4

. l

..-.-_- - - -.---. - ~.. _..-- -- - -

Ch0Ptor 48 RESULTS PAGE 19 I.

l 1.30 I

l.25-t r

M

! i.i -

. :. l :*!.

I 3

g i.00

..,. 7,y..

8.*

I l

0*'5-C b.

t')*.,.,***g, g.

%g O

+

I e

0. 0-o r

s.

. ****.*I*.

1 g l, e.

' g. *,* j ;.,

l i

O 0.00-

' I O.15-f

-01 0.0 01 02 0.3 0.4 HOT CHANNEL OUILET 90RLl1Y o

Figure 4.84 - Measured to Predicted Critical Neat Flus Ratie vs. Not Channel Outlet tuality I

I

g ene,so inzutr.

ca.:

20 l

I s

I 1.30 i.25U Ll 204 i

~

i i.il-c g

0

$. p. *.:.* ** **{'*.**,t +

I **

- e.**g,*4..

, of..

s' 5-

g

t.f g*e, s..:%so*.

\\

,.0,.

u.*

k*'d Id.

g,g,

.*v..**;.c.} n.f*345 s%

0.95-l

,3

0. 0

<s.. *.,jt. 4. ;

.t g,**

**,3.

s I

i 0

0.80-0.45 0.40 i

350-400 450 500 550 600 650 INLET 1E,PERATURE-tDEOREE6 Fi

+

i 1i Figure 4.5 - Measured to Fredicted Critical Neat Flus Ratie vs. Inlet Temperature ('F)

E.

E E.

j

)

Chaptor 4:

RESULTS Pact 31 1

I l

rntoutwCv 24o

22. -

I

,h too -

l

! ! !i I

i

i. -

l l

140 -

l i

i i

l l

j j

l l

l Bo -

l l

l i

i l

l l l

I l! l

"~

! l! ! ! b l l l i

l a-l l

l l

l h

"~

l l

i:i:::i:i l

l ll l

l

i:::!::::

l l

oNo

'5'.'t o c.es o.so c.ss i co i.os i.io i.is MEA 80Rio to PRtolCffo CHF RAflo I

Figure 4.6 - Mistogram of Measured-to-Predicted CNF Ratios I

I

)

Chaptor 48 RESULTS PAGE 22 I

TABLE 4.1 l

MEASURED-TO-PREDICTED CRITICAL HEAT FLUX RATIO STATISTICS BY AXIAL POWER PROFILE

]

Axial 4 of M/F Standard Profile obrervations Mean Deviation i

Unifora 345 1.002 0.0441 Cosine 176 1.011 0.0441 J

u* sin (u) 344 0.995 0.0429 Total 945

-1.001 0.0434 FABLE 4.2 I

MEASURED-TO-PREDICTED-CR::TICAL NEAT FLUX RATIO STATISTICS BY S'IBCitANNEL CELL TYPE l

Cell' 4 ef M/P Standard i

Type Observations Mean Deviation Thimble 207 0.997 0.0432 TrPical 734 1.001 0.0439 Total 945 1.001 0.0434 TABLE 4.3 I

MEASURED-TO-PREDICTED CRITICAL HEAT FLUX RATIO STATISTICS j

BT ROD DIAMETER (INCH)

Rod-U" 4 of M/P Standard Diametee Observations Mean

' Deviation I

0.374*'

449 1.013 0.0434 0.428' 456 0.944 0.0423 I

Total 945 1.001 0.0434 l

-I I

j

-l I

s ChaPtor,4:

RESULTS PAGE 23 lI ll TABLE 4.4 MEASURED-70-PREDICTED CRITICAL HEAT FLUX RATIO STLT7STICS I

BY TEST SECTION HEATED LENGTH (FT)

Heated 4 of M/P Standard Length Observations Mean Deviation 8 ft 335 1.003 0.0846 14 ft 610 1.000 0.0834 Total 945 1.001 0.0834 TABLE 4.5 l

I MEASURED-TO-PREDICTED CRITICAL HEAT FLUX RATIO STATISTICS BY GRID SPACIMG (IMCHES)

Grid 8 of M/F Standard SPacinS Observations Nean Deviation 13" 35 0.954 0.0716 20" 101 0.987 0.0857 12" 140 1.011 0.0835 26' 424 1.003 0.0434 32*

105 0.99a 0.0416 Total 945 1.001 0.0434.

I G.

i p

t

I ChcPter 48 RESULTS PAGE 24

_5k

sm t

i i

TABLE 4.6 COMPARISON or

SUMMARY

STATISTICS FROM VIRGIMIA POWER, WESTIMGHOUSS AND MRC CALCULATIONS l

l EPRI M/F Mumber Test Analyst Code M/F Standard of Section+

Mean Deviation Points L

1 161 (A-1)

VA Power COBRA IIIo/MIT 1.0064 0.0630 71 t

l Westinghouse THIMC 0.9964 0.0655 71 MRC COBRA III 0.9912 0.0640 71 MRC COBRA IV 0.9910 0.0640 71 s

160 (A-3)

VA Power COBRA IIIo/MIT 1.0079 0.0904 65 Westinghouse THIMC 1.0502 0.1080 67 MRC COBRA IV 1.0360 0.0938 67 I

+The corresponding Westinghouse label is listed in Parentheses.

I t

1 J

I ChePter 48 RESULTS PAGE 25 j

I TABLE 4.7 MEASURED-70-PREDICTED CRITICAL HEAT FLUX RATIO

SUMMARY

STATISTICS POR ALL TEST SECTIONS EPRI Number M/P Axial Rod Grid Heated Test of M/P Standard Power Diameter Spacing Length Section+

Points Mean Deviation Profile (inch)

(inches)

(ft) i 114 (A-6) 32 0.9914 0.0524 Cosine 0.412 to 4

125 (A-7) 33 0.9292 0.0544 u*sintu) 0.422 to 4

117 (A-4) 36 1.0044 0.0492 u* sin (u) 0.422 16 4

i 131 (A-9) 32 1.0233 0.0784 u* sin (u) 0.422 26 14 132 (A-10) 36 1.0355 0.0944 u* sin (u) 0.422 to 14 133 (A-11) 35 0.9539 0.0716 u* sin (u) 0.422 13 14-134 ( A - 1 ;, )

34 1.0044 0.0402 u* sin (u) 0.481 32 14

{

I 139 (A-17) 37 0.9705 0.0473 u* win (u) 0.423 32 14 140 (A-13) 30 1.0057 0.0712 u*sintu) 0.483 38 4

u sin (u) 0.428 to 14 146 (A-16) 37 1.0003 0.0576 s

144'(A-14) 70 1.0035 0.0786' u* sin (u) 0.422 16 14 153 (A-15) 40 0.9149 0.0554 Uniform 0.482 26 14 1

156 (A-2) 70 1.0116 0.0700 Uniform 0.374 26 14 5

t 157 (A-4) 76 1.0123 0.0744 Uniform 0.374 26 4

y 154 (A-19) 63 1.0854 0.0941 Uniform 0.374 to 4

160 (A-3) 65 1.bd79 0.0904 Uniform 0.374 it 4

161 (A-1) 71 1.0064 0.0630 Unifera 0.374 it 14 162 (A-14) 70 0.9450 0.0757 Cosine 0.374 28 14 164 (A-5) 74 1.0443 0.0917 Cosine 0.374 tt 14

. ~...

Total 945 1.0010 0.0434

+ Westinghouse labels in Parentheses.

P.

. ~

1 I

Chaptor 5:

App 11 CATION OF THE WRB-1 CORRE1ATION PAGE 26 j

i 5.0 Application of the WRB-1 Correlation i

Westinghouse has been using the WRB-1 correlation, along with I

the 1.17 THINC/WRB-1 limite on 17x17 standard fuel since 1978.

In addition.

Westinghouse presented data in Reference 4 which showed that the 1.17 DNBR limit provided the necessary 95/95 protection for the 14x14 and 17x17 OTA products, and demonstrated that the 4

limit was also applicable to the 15x15 OTA.

The extension of the

!1 WRB-1 limit to the CFA product was subsequently approved' by the' NRC.

i Virginia power's ir. tended application of de WRB-1 corrslation will be for analysen of the Westinghouse 17x17 standard 'R' prid

]

fuel at North Anna and the Westinghouse 15x15 0FA-type + fuel at surry.

Because no "L"

grid data were included in the test r.opulation, the WEB-1 o9trelation will not be applied to 15x15 standard fuel, t

Even though only "R'

prid data have been examined in this

report, Virginia power has concluded that the use of the be extended to West'inghouse OFA-type fuel.

COBRA / WEB-1 limit een

'I l

COBRA /WRB-1 and TNIMC/WRS-1 have been shown to perform comparably by the analyses doeunented herein.

The similarity of the COBRA and 1

+Beginning with the cycle 10 reloads. Virginia power plans to use a fuel product at surry which includes additional features such as a

a renovable top nossle, but is geometrically identical (including i

the gries) to the Westinghouse 15x15 0FA in the active core 1

region.

f

Choptor 58 ' APPLICATION OF THE WRS-1 CORRELATION PAGE 27 i

THINC codes is well establisheds Reference 3 was approved in part on the basis of an excellent comparison between the two codes.

As a

result, the *R* prid C08RA/WKB-1 DNBR limit may also be applied to the Westinghouse OTA-type products.

This application would be consistent with the licensed use of THINC/WRB-1.

I COBRA /WR8al will be applied over the saae range of conditions as THINC/WRB-1.

1 I

I I

I i

l

!t l

y chopter 6:

CONCLUSIONS PAGE 24 I

l 6.0 conclusions The WRB-1 CHF correlation has been qualified in Virginia power's thermal-hydraulics code COBRA IIIc/MIT.

Data reduction has shown that the use of a

1.17 DMBR limit provides 952 non-DMS probability at a

954 confidence level.

These results are applicable to Westinghouse 15x15 and 17x17 "R' grid fuel and to Westinghouse 14x14, 15x15 and 17x17 0FA-type fuel.

1 I

I I

I.

I 8

I I

I

PAGE 19 I

E REFERENCES I

1.

Mottky.

F.

E.,

et al.

"New Westinghouse Correlation WRB-1 for Predicting Critical Heat flux in Rod Bundles uith Mixing Vane i

Grids."

WCAP-4762-P-A (proprietary) and WCAP-4763-A (non+ proprietary) (July, 1984).

I 1.

511a.

F.

W.

and M.

L.

Basehores "Vepco Reactor Core Thermal-Mydraulie Analysis Using the COBRA IIIC/MIT Computer Code." VEP-TRD-33-A (October, 1983).

3.

righetti.

C.

F.

and D.

G.

Reddy

" Parametric Study of CHF Data." EPRI MP-1609. Vol.

3.

Part 1 (September. 1983).

I 4.

Rosal.

E.

R.,

et al.

"High Pressure Rod Bundle DMR Data with Axially Monunifera Heat Flux." Muclear Engineering and Design.

Vol. 31 (1974).

5.

Anderson.

R.

C.

" Statistical DMBR Evaluation Methodology."

VEP-ME-t (July, 1985).

I 6.' Letter from J.

F.

Stols (MRC) to C.

Eicheldinger (Westinghouse).

" Staff Evaluation of WCAP-7956.

WCAP-4054.

WCAP-8567.

and WCAP-4748 " dated April 19. 1978.

7.

" Assessment of the Assumption of Moraality (Employing Individual observed values)." ANSI M15.15 (1974).

8.

Letter from E.

P.

Rahe (Westinghouse) to C.

O.

Thomas (MRC).

" Basis for the Applicability of the WR8-1 CHF Correlation to 15x15 ora and 14x14 0rA Fuel.

Supplement

1. - WCAP-4762 I

(Proprietary)"

and WCAP-4763 (Mon-Proprietary). dated November it. 1988.

I 9.

Letter from C.

O. Thomas (MRC) to E.

P.

Rahe (Westinghouse).

"Acceptamos for Referencing of Licensing Topical Report WCAP-8798(P)/WCAP-4763(MP).

Supplement 1.

' Basis' for the Applicability.of the WRB-1 Correlation to 15x15 ora and 14x14 0FR Fuel, dated June 89. 1984.

I 9

I I

I

l PAGE 30 l

t i

I 8

I APPENDIX A DATA LISTING I

t I

l' I

I 8

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A-14 389 1220 1.032 0.96099 0.568 403.9 1510 3.473 0.093 0.6644 4

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390 1221 1.000 1.00000 0.613 487.5 1810 3.492 0.067 0.6952 4

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396 1227 1.058 0.94518 0.478 400.4 1500 1.998 0.106 0.5733 4

A-14 397 1228 1.106 0.90416 0.305 542.6 1815 1.965 0.198 0.3825 4

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A-14 403 1235 1.031 0.96993 0.369 496.9 2405 1.543 0.130 0.4314 4

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A-14 408 1240 1.020 0.90039 0.280 563.5 1505 1.502 0.315 0.3238 4

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. STATISTICAL AN6 LYSIS OF COBRN-1 OleR DATA 15:17 THURSOAY, OCTOBEi! 30, 2986 - 23 TSTSECTN=157 - "

OSS RLM 9999R

^W HEATFLLDC TIN PRESSURE MASSFLts(-

EXITQUAL CHF ITERATHS EST 567 1601 0.951 1.05152 0.588'

'516.7 1515 2.501 0.137 0.6208 4

A-04 568 1602 1.030 0.97087

- *8.628 515.4 1500 2.989

.0.103 0.7182 4

A-04 3'

569 1603 1.0E0 0.98039 0.697 514.8.

1515 3.492 0.084 0.7898 4

A-04 570 1604 0.828 1.20773-

.0.434 576.2 1505 2.017 0.246 0.3992 10 A-04

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571 1605 0.976 1.02459

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566.8 1505 2.481 0.184 0.5018 3

A-04 572 1606 0.954 1.94822 t 0.523 564.5 1505 3.010 P.161 0.5559' 3

~ A-05 4-09 573 1607 0.946 1.05704 0.542' 571.7 1505 3.477 0,153 0.5696 3

574 1608 0.928 1.07759 0.702 539.5 1800 3.446.

0.0P +

0.7237 4

A-0*

575 16091 0.924

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A-04 576 1610 0.960

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. -* 0.697 4 71.0 1505 2.629 0.084 0.7435 4

A-04 577 1611 1.609 9.99108 0.377 570.0 1795 1.531 0.231 0.4227 4

A-04

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578 1612

- 1.048 9.95429 i9.471 549.9 1805 2.0 73 0.149 0.5480 4

A-04 579 1613 1.000

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.0.503 563.4' 2005 2.515 0.143 0.5536 4

A

580 1614 0.922

-1.00440 ~-9.942 574.6 1815 2.963 0.143 0.5549 4

A-04 581 1615 1.040 e.96134

.e.60s 561.9 1805 3.514 0.096.

0.7023 4

A-04 4

582 1616 1.009 G.99198 ~l 9.382 517.1 1805 1.067 0.269 0.4283 4

A-04 583 1617 0.951 1.05152

. 0.M4 516.1 1800 1.517 0.193 0.4942 4

A-04' 584 1618 0.925 1.0010e 8.416 495.3

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A-04 585 1619 0.997 1.00301 0.444 513.0 1505 1.462 0.225 0.4919 4

A-04

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586 1620 1.209 0.82713 S.445 393.6 1790 1.035 0.111 0.5976 5

A-04

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587 1621 0.9M 1.03520 0.617 453.8 1495 1.989 0.116 0.6623 5

A-04 548 1625 0.948 1.85485 0.679 427.9 1800 1.957 0.056 0.71 %

5 A-04 f

589 1624 0.912 1.09M9 0.518 387.7 1510 1.051 0.238 0.5246 5

A-04 590 1625 0.871 1.14811 8.623 402.9 1505 1.526 0.151 0.6025 5

A-04 i

591 1626 0.857 1.1486 0.719 410.7 1505 1.967 0.102 0.6841 5

A-04 592 1627 9.969 1.03199 0.439 575.9 1495

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A-04 3'

i 594 1629 0.983 1.01729 0.436

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A-04 595 1630 0.900 1.11111 0.551 457.7 1500 1.538 0.189 0.5510 4

A-04 596

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597 1632 1.088

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A-04 598 1633 0.987 1.01317 8.498 396.8 2105 1.057 0.122 0.5456 5

A-04 599 1634 0.949 1.05374 0.518 392.3 2425 1.949 0.079 0.5458 5

A-04 600 1635 0.921 1.08579 9.653 398.4 1405 1.505

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A-04 601 1636 0.914 1.09409 0.635 400.7 2115 1.498 0.048

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A-04

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A-04 4:

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TSTSECTN2158 i

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TIN PRESStatt.

MASSFLtBC EXIIGUAL CHF ITERATHS MST

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l 603 1638 0.958 1.04384 0.517 555.1 2100 2.035 0.138 0.5527 4

A-19 604 1639 1.010 0.99010 0.571 554.7 2105 2.503 0.092 0.6433 4

A-19 E

605 IMO 0.971 1.02987 0.647 556.4 2100 3.007 0.077 0.7012 4

A-19 606 1641 0.910 1.09890 9.738 556.4 2105 3.496 0.069 0.7491 4

A-19 i

607 IM2

0. 9M 1.03734 0.558 552.2 2395 2.048 0.097 0.5999 4

A-19 608 1643 9.990 1.01010 0.643 547.6 2425 2.524 0.039 0.7101 4

A-19 609 IM4 0.979 1.02145 0.695 558.0-2395 2.996 9.037 0.7591 4

A-19 610 1M5 0.9M 1.03734 0.767 562.9 2435 3.446 0.022 0.8251 4

A-19 611 IM7 1.188 0.84175 0.444 611.7 2400 2.496 0.117 0.5087 4

A-19 C

612 IM8 1.068 0.93633 0.519 613.7

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6 December 19, 1988 1

5 i

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U. S. Nuclear Regulatory Commiss.on Serial No.

88-833=

Attn: Document Control-Desk N0/KLB:R1 Washington. D.C.

20555 Docket Nos.

50-338 50-339~

50-280.

l 50-281 License Nos.

NPF !

NPF-7.

l B-DPR-32 DPR Gentlemen:

VIRGINIA ~ ELECTRIC AND POWER COMPANY NOMIH ANNA POWER 5TATIOR UNIT 5 1 AND 2 i

I

~

SURRY POWER STATION UNIT 51 Ml"J 2 TOPICAL REPORT VEP-NE-3 "QUALIFICATE0N OF THE WRB-1 CHF CORRELATION IN THE VIR63NIA POWER COBRA CODE" Several questions were informally raised by your staff concerning the subject topical report during a meeting on_ December 8,

1988.

We have prepared I

responses-to those questions and have included them as Attachment 1.

1 Very truly yours, i

,W. R. Cartwright Attachment l

cc: U. S. Nuclear Regulatory Commission I

101 Marietta Street, N. W.

Suite:2900 Atlanta. GA 30323.

Mr. J. L. Caldwell i

NRC Senior Resident Inspector North Anna Power Station

~

Mr. W. Holland NRC Senior Resident Inspector Surry Power Station I

g -

l h;h c

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1

.I lu Attactment Response to Infonnel Questions Asked by the Mtc I

on Topical-Report VEP-NE-3

" Qualification of the WRB-1 CHF Correlation in the Virginia Power CCBRA Code" I

I I

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Pleas'e> provide' a copy of Ref.

2,- VEP-FRO-33-A. -for.

j 1

s Question No..:1:-

details ~ on the thermal-hycraulic lmodeling used in COBRA-!!IC/MIT; irf' g

evaluating' data for the WRB-1 correlation.

t l

t-3 Answer:~ Attached.

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i Q. uestion No. 2: ' identify the 22 data points eliminated from consideration because ' they " differed by -more than ten standard deviations."

What g

quantity was c:meared to determine this difference? If these data points-i,g were included in the-THINC/WRS-1 data base, they should also be good for 1

the data base with the COBRA-IIIC/MIT code.

Explain what is wrong with 1

these data points that would justify excluding them.

Answer:-

The entire COBRA /WRB-1 database was taken from EPRI report NP-2509.

The data are also listed in Westinghouse proprietary report

, 3 WCAP-8762-P-A.

Some minor variations exist between the reported bundle 4

l3 conditions in the two data sets because Colust.ia University reduced the data for EPRI-NP-2609, whereas Westinghouse reduced the-dat'a themselves i

for WCAP-8762-P-A.

.As a result,.it was 'not -' uncommon to see inlet' temperatures differ by a few tenths of a degree, for example.

Similar t

variations in average heat flux, flow and pressure also existed.

When*

l the difference -in. one of the reported bundle conditions (either ER temperature, heat flux or flow) exceeded.- the average difference by ten-

,5 standard deviations, the point was deleted from the C08RA/WRB-1 database 1

- as being a-possible. typograpnical error.

The deleted points are listed-4 below.

I

[

Test Section WCAP-8762-P-A Number Run Numoer-Appendix Deleted Number ( s)

'I 124 '

A-6 1

479 i

131' A-9 3

621, 637, 638 a

'g

'133 A-11 3

702, 706, 707 140 A-13 1

887 146-A-16 2-1110, 1128 cE 148 A-14 1

1182 5

153 A-15 2

1396, 1401-i 156 A-2 3

1459, 1462 1471 l3, 157 A-4 2

1586, 1596-

3; 158 A-19 5

1646, 1661, 1674, 1677, 1696 160 A-3 2

1725,.1771A Total 25 The statement that 22 data points were deleted, rather than 25, was a:

e

. typographical error, g

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Question No. 3:

Please identify more fully the contents of the cata l

tables in Appendix A I

I, is'the value under the heading " heat flux" the average heat flux in e'

the test section for that point?

is the " mass flux" the inlet mass flux for the bundle, or the local' e

mass flux calculated in the hot enannel at the location of the MONBR?

is the value given for the " exit quality" the bundle average exit e

quality?

I 1

is the value under the heading "CHF" the predicted critical heat flux e

at the location of MON 8R7 what is the meaning of the column labeled " iterations"?. Does this3 represent the numoer of iterations required in 'the COBRA-IIIC/MIT coce to obtain tho' thermal-hydraulic solution for the conditions of-each dats-point?

-t does the axial location of the MONBR have any relation to the location e

' where the CHF was actually measured for a given data point?

Answers:

HEATFLUX is the measured bundle-average heat flux.

e MASSFLUX is the measured bundle-average mass flux.

EXITQUAL is the predicted hot-channel outlet quality.

CHF is predicted critical heat flux at the pcInt of alnimus DN8R.

a I

ITERATNS is the number of hydraulic iterations required by COBRA to e

resolve the flow field.

Flowfield turbulence induced by the etking vane grids suppresses CHF but decays with distance, so that CHF generally occurs just upstream of a saixing vane grid (or at the outlet lof uniformly exially heated bundles), as reflected in the EPRI-NP-2609. database.

The WRB-1 I

correlation models this turbulence, thus predicting CHF as the'same axial location where it was first observed for most of the data.

In some cases, WRS-1 predicts the minimum DNOR just upstrees of. another grid. WCAP-8782-P-A reported (on page 3-6) that WRS-1 predicted the location of CHF with an accuracy of 81.75.

A review of the axial positions of C08AANRS-l's-minimum DNOR prediction showed that-its 3

accuracy was comparable to that of THINCNRB-1, as expected.

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- diskettei.c other magnetic media so that we can perfera some independent w

Questioni % 4: Please provide a copy of the data in Appendix A on floppy statisticsl evaluations of the WRS correlatiots-- behavior -with COBRA-!!!C/MIT.

Answer:

Those data have already been provided in the January 29, 1987 I

- package in which VEP-NE-3 was submitted for NRC review I

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I Question No. 5:

State erplicitly the intended range. of. application of _

the WRB-1 correlation with CCBRA-IIIC/MIT, including operating conditions I.

and fuel designs.

The statement that' it is to be "the same as WRB-1/THINC" is helpful, but by itself is too general.

Answer:

1440 5 pressure 5 2490 psia 0.9 s Mass Flux 13.7 Mlba/hr-fta_

Local Quality s 0.30 i

As a result of concerns raised by Question No. 6. COBRANRS-1 will not be used when the local heat flux exceeds _1.0 MBTU/hr-fts, or for fuel with -

i 13" mixing vane grid spacing.

As noted in Section 5.0 (paga 26) of VEP-NE-3, C08RANR8-1 will be used 1

for analysis of the Westinghouse 17x17 standard "R" grid fuel and the I

Westinghouse 17x17 Vantage 5H zircaloy grid fuel at North Anna and' the-Westinghouse 15x15 0FA-type fuel at Surry.

COBRAARS-1 will not be-applied to the 15x15 "L" grid-fuel product.

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- Question No.26:~ There are several indications of nonconservative behavior in the analyses presented.

g (a) Figure 4.1 :seems to indicate that the WR8-1/ COBRA combination is nonconservati_ve for larger critical heat fluxes.

(b) The' dif ference in rod diameters in Table 4.3 is statistically significant, with an apparent nonconservatism for 0.422" diameter rods.

(c)

The results in Table 4.5,suggest that the correlation is nonconservative for 13" and 20" grid spacings.

Jg Please comment on each of the above observations, and address whether the 3

apparent' nonconservatisms are related to the' variables noted or whether they.are due to interactions of these with other variables.

Answers:

(a) CHF is overpredicted above 1.0 M870/hr-f t.-' However, C08RA/WR8-1 can.

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only be used =at heat fluxes up to approximately 0.8 M8TU/hr-fta,' based

(;

upon the following:

t q

Average (uprated) heat flux at North Anna and Surry < 0.21' MBTU/hr-ft8 Overpower limit = 1185 1

Design radial factor < 1.65 for both plants Axial factor s 2 for both plants L

0.21

1.18

1.65

2 = 0.82 MBTU/hr-ft8 L

Since-Virginia Power chose to qualify COBRA /WRB-1 with the same "R" grid hI database over which THINC/WRS-1 was qualified, these data were included in VEP-NE-3.

However, C08AAMIB-1 will not be. applied above 1.0 MBTU/hr-ft* for Virginia Power's reactors.

L (b) An examination of the summary statistics on page 25 of VEP-NE-3 show that the -slightly low 0.422* M/P mean is-due primarily to the contributions of Tese Sections #125 (mean = 0.93). #153-(mean = 0.915)

I and #133 (mean = 0.95).

TS #125 consists wholly of-high heat flux data where the Surry and North Anna plants cannot operate. They were included q

for completeness, however, in order to be consistent with the philosophy of demonstrating the similarity.of COBRAMtB-1. and THINCMtS-1.. The TS

153 data, although found with a low mean, matched their THINC/WRB-1 counterpart well (mean = 0.93, from page 3-2 of WCAP-8762-P-A).. The' TS

133 COMAMIB4 data also differed only slightly from their THINCMtB-1 I

counterparts (M/P acans of 0.95 and 0.97 respectively). Further, as will be noted below, the TS #133 data were not needed, but were included to preserve the integrity of the. original THINCMl8-1 "R" grid database as much as possible. ' The 0.422" data mean without TS #125 and TS #133 -is 0.996, nearly ' identical to the '1.001 average of-the full dataset (945 points).

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I-(c) The:13" grid' spacing data (T5 #133) were included for completeness.

although Virginia Power has no 13" grid spacing fuel; the grid. spacing -

is 20" for our 17x17 fuel and 26" for our 15x15 fuel.

The slightly low average for the 20" grid spacing data is due entirely to TS #125. which:

i consists entirely of high heat flux data where Surry and North Anna cannot operate.

The 1.17 limit is appropriate for the application of COBRAARB-1 to Surry-(26" grids) and North Anna (20" grids) at the conditions listed in the response to Question No. 5.

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The following table is prou ded as additional evidence of the similarity of COBRAARB-1 and THINC-WRB-1.

l Test Section M/P Mean M/P Standard Deviation Data Points I

(VPN)

(VPN) 161 (A-1) 1.0064/0.9964 0.0630/0.0655 71/71 H

156(A-2) 1.0116/1.0041_

0.0780/0.0805 70/73 B

160 (A-3) 1.0079/1.0502 0.0908/0.1020 65/67 157 (A-4) 1.0123/1.0136 0.0788/0.0848 76/78-164-(A-5) 1.0443/1.0022 0.0917/0.0852 74/74-

_ sa 124 ( A-6) 0.9918/1.0042 0.0528/0.0528 32/33 i

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125(A-7) 0.9292/0.9937 0.0588/0.0649 33/33 127_(A-8) 1.0084/0.9846 0.0892/0.0922 36/36 131 (A-9) 1.0233/1.0584 0.0788/0.0816 32/35 I

132(A-10) 1.0355/1.0100 0.0988/0.0915 36/36 133 (A-11) 0.9539/0.9737 0.0716/0.0781 35/38 134 ( A-12) 1.0088/1.0238 0.0802/0.0752 38/38 140 (A-13)

-1.0057/0.9913 0.0712/0.0724 30/31 5

4 148 (A-14) 1.0035/1.0466

'O.07e6/0.0829 70/71 l

153 ( A-15) 0.9149/0.9321 0.0558/0.0595 40/42 i

146 (A-16) 1.0003/1.0141 0.0576/0.0579 37/39 I

139 (A-17) 0.9706/0.9728 0.0873/0.0887 37/37 162 (A-18) 0.9850/1.0002 0.0757/0.0796 70/70 158 (A-19) 1.0254/1.0303 0.0941/0.1048 63/68 i

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Question No - 7:

The investigations presented in Tables.4.1-4.5 and-L Figures 4~.2-4.5 only look at the effects on M/P of the variables one-at-a-time, and thus are only useful in detecting gross biases 'in L

correlation performance.

It is also important to employ techniques that 3

look. at the variables several-at-a-time so as to provide for detecting og biases that may occur for various combinations of levels of the variables.

g

' By " variables" we nr.an the correlation's independent variables or-other variables -that may have a potential effect (e.g., bundle differences).

An ~ exploratory technique for investigating the effects of several-I

-variables at a. time is to compute means and standard deviations for groups of M/P values formed by taking the variables two-at-a-time, i

three-at-a-time, and so on.

Statistical techniques'such as multiple L3 regression analysis or seve"l-factor analysis of variance can be used 3

to supplement the exploratory investigation. of the ef fects of othe variables.

Were any investigations along these lines performed, and if.

so, what were the results?

I

Answer:

Yes.

Extensive Analysis of Variance (ANOVA)' was performed on the entire WRS-1 database when the NRC first reviewed the correlation as.

I, presented in WCAP-8762-P-A. The ANOVA results are discussed in the NRC's r

Safety-' Evaluation of WRB-1 (letter frca J.. F. - S tolz, NRC, to - C. -

Eicheidinger, Westinghouse, " Staff Evaluation of WCAP-7956, WCAP-8054.

WCAP-8567 and WCAP-8762," dated April 19, 1978).on pages 8-13 and Table I

V.

Some variation in the subgroup statistics was identified, and the NRC refused to authorize the 1.17 limit for "L" grid fuel.

The 1.17 limit was subsequently approved for the remainder of the dataset (the "R" grid i

s ~

fuel).

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'4 Question No. 8:

It is'a well-established fact that a correlation' derived by -least-squares ' methods will tend to fit the data it was derived from better than "new" data,' even when the "new" data is from the same range I

of conditions. Therefore, it is important to validate a correlation using-data that were not employed in its derivation.

If it can be determined that the correlation actually does perfore equally well, or nearly as well, ifor, "new" data as for the data it was developed from, then its performance can be evaluatud using both the "new" data and the development data. However, it should not be assumed, a priori, that this is the case, Was this assumption =-tested in evaluating ~ the WRS-1 correlation's I

performance with COBRA-!!!c/MIT? If not -why not?

If so, how, and what were the results?

I Answer:'. This assumption was not tested with COBRA, The need for such validation is sitigated for large databases when the correlation is to -

be applied over a narrow range of conditions and fuel types, as is the cast with WR8-1.

The adequacy of the data density is illustrated by.

j VEP-NE-3 Figures ' 4.1-4.5, for enaaple.

WCAP-8762-P-A does not state whether the correlation was tested with "new" data, 1

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'I Question No. 9: lTheONBRlimitformula the usual 95/95 tolerance interval form. given in Equation 4.1-is' not of Equation 4.1 contains the term 5

1.645'k instead~of the usual K,-values of which are: tabulated (e.g...in i

I exists.(graph by.D. B. Owen). An approximation for K of the form 1.645'k the mono which was popular many years-ago before K was tabulated), but it is nonconservative in many cases. Describe how k is defined, and explain why the 1.645*k multiplier is used instead of the standard K.

I Answer:

The tolerance limit accounts for the natural randoeness of a variable, as well as the uncertainty in its mean and standard deviation.

'E Randonness is accounted for, with 955 probability, by the 1.645-5 multiplier; the k-coefficient accounts for the two small-sample effects, although only one of them is mentioned in VEP-NE-3. The product 1.645'k is thus identical to the Owen's factor. for enaaple, and both methods I

yield a 1.17 DNBA limit for the COBRA ARS-1 database.

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- Question No.10:

The ONBR liett of.1.17 for WRB-1 with COBRA-IIIc/MIT:

was obtained using the mean and standard deviation of the entire data base of 945 points.. This procedure assumes that the M/P values come from a' I-common-population _ and that the 945 points are a random sample of the conditions and assemblies that the correlation is appitcable to. The data are not a randoa sample and it is unlikely (based on statistical theory.

I our experience with other CHF correlations); that. the correlation and performs equally well over its full range of applicability. The " common 4

population"'assumpt 4n should never be taken as a given assumption for a j

correlation of this type.

Discuss any investigations you performed to B

check the common population assumption or to identify different-subpopulations ' (a).

If you did not perform any such investigations, discuss why not.

Answer:

Virginia Power performed no investigation of the " common-*

population" assumption.

This assumption was investigated thorwghly by-the NRC in their original review of WCAP-8762-P-A and reported on in their

{

safety evaluation (letter from J. F. Stoir, NRC, to C.

Eiche1dinger, Westinghouse; dated April 19, 1978).

As a result of ' thatJ review,_

'l Westinghouse was not allowed to use the 1.17 limit on "L" grid fuel.-

R (Virginia Power has not requested approval for the use of WRB-1 withL"L" d

grid fuel.)

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y (a) The normality of. the M/P values from the whole data base,'as

. demonstrated by the O' test, is not relevant to testing the " common i

population" assumption.

Data from several different normel populations I

can be normally distributed when combined, and_ thus normality of the M/P values doesn't " prove" that they came from a common population.- Further, it is not necessary to have 'all of the M/P values cose from a commen normally distributed population in order to apply normal theory' tolerance-I j

interval formulas to calculate ONBR limits. c All that is required is that the set of potential M/P values for any set of conditions be normally 1

distributed. The individual normal distribution's may be different (i.e.,

1 have different means and standard' deviations).

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ML20059L076

Text

Dear Mr. Cartwright:

SUBJECT:

Enclosure:

1.0 INTRODUCTION

3.0 CONCLUSION

4.0 REFERENCES

SUMMARY

SUMMARY

=

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