ML20108D209

Nonproprietary App C, Mcguire/Catawba Plant Specific Data Mark-BW Fuel BWU-Z CHF Correlation to DPC Thermal- Hydraulic Statistical Core Design Methodology

ML20108D209
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	04/30/1996
From:	DUKE POWER CO.
To:
Shared Package
ML19355D094	List: ML20108D195 ML20108D209
References
DPC-NE-2005P-A, DPC-NE-2005P-A-APP-C, NUDOCS 9605070326
Download: ML20108D209 (20)

Text

-

DPC-NE-2005P-A Duke Power Company Thermal-Hydraulic Statisticel Core Design Methodology APPENDIX C McGuire/ Catawba Plant Specific Data Mark-BW Fuel BWU-Z CHF Correlation April 1996 i

I 9605070326 960426 PDR ADOCK 05000369 P

PDR

4 This Appendix contains the plant specific data and limits for the McGuire and Catawba Nuclear Stations with Mark-BW fuel using the BWT7-Z form of the BWU critical heat flux correlation.

The thermal hydraulic statistical! core design analysis was performed as described in the main body of this report.

Plant Soecific Data This analysis is for the McGuire and Catawba plants (four loop Westinghouse PNR's) with Mark-BW fuel assemblies as described in Reference C-1.

The parameter uncertainties and statepoint ranges were selected to bound the unit and cycle specific values of the McGuire and Catawba stations.

Thermal Hydraulic Code and Model The VIPRE-01 thermal-hydraulic computer code described in Reference C-3 and the McGuire/ Catawba eight channel code model approved in Reference C-1 are used in this analysis, c-1

i F

l 4

1

-Critical Heat Flux Correlation i

l The BWU-Z form of the BNU critical heat flux correlation described l

j

ba Reference' C-2 is used for all statepoint analyses.

This correlation was developed by BNFC for application to the Mark-BW. fuel design.

t j

Reference C-2 was performed with the LYNXT thermal-hydraulic computer codes.

The correlation was programmed into the VIPRE-01 thermal-hydraulic computer code by Duke Power Company and the BWU-Z CHF data base analyzed in its entirety.

The results-of this analysis are shown in Table C-1.

The resulting Average M/P value and data standard l

deviation are within 1% of the values reported in Reference C-2.

l Figures C-1 through C-5 graphically show the results of this 1

evaluation.

Figure C-1 shows there is no bias of measured CHF values to VIPRE-01 predicted values for the data base.

Figure C-2 shows a histogram of the VIPRE-01 M/P ratios for the 530 point data base.

l L

Figures C-3 through C-5 show there is no bias with the VIPRE-01 calculated M/P ratios with respect to mass velocity, pressure, or thermodynamic quality.

These figures compare closely with the same parameter representations in Reference C-2.

l Based on the results shown in Table C-1 and Figures C-1 through C-5, the BWU-Z form of the BWU CHF correlation licensed in Reference C-2 can be used'in DNBR calculations with VIPRE-01 for Mark-BW fuel.

l' i

C-2 l

. =.

l Statenoints The statepoint conditions evaluated in this analysis are listed in Table C-2.

These statepoints represent the range of conditions to l

which the statistical DNB analyses limit will be applied.

Kev Parameters and Uncertainties The key parameters and their uncertainty magnitude and associated distribution used in this analysis are listed on Table C-2.

The uncertainties were selected to bound the values calculated for each parameter at McGuire and Catawba.

The resulting range of key parameter values generated in this analyses is listed on Table C-5.

DNB Statistical Desian Limits The statistical design limit for each statepoint evaluated is listed on Table C-4.

Section 1 of Table C-4 contains the 500 case runs and Section 2 contains the 5000 case runs.

The number of cases was increased from 3000 to 5000 as described in Attachment 1 of the main body of the report.

All statepoint SDL values listed in this analysis are normally distributed.

The maximum statepoint statistical DNBR value in Table C-4 for the 5000 case propagations was 1.364.

C-3

Therefore, the statistical design limit using the BWU-Z form of the BWU CHF correlation for Mark-BW fuel at McGuire/ Catawba was conservatively determined to be [

].

t C-4

FIGURE C-1 Measured CHF Versus Predicted CHF Mark-BW Data Base i

l

!.2 O

T O

O y

l 0.9 -

0 E o U

th J

s u

O e

O 00 i

d cu

.ii

'0 0

0%

I

Ogr, a

0. 6_

O O

a y

0 y

t u

(

8,P'f.J r

r

{ 0.5-0 l

2

%D 0.0 i

0.0 0.3 0.6 0.9 1.2 Predicted CHf Mblu/hr-Il2 i

C-5

1 FIGURE C-2 Distribution of CHF Ratios l

l Mark-BW Data Base 1

l FREQUENCY '

120-110:

100:

i 90-1 80:

70:

l I

60:

l 50-40-i l

30-20-10:

I i

O i

I l'

0.b5 0.70 0.75 0.80 0.85 0.90 0.95 1.00 1.05 1.10 1.15 1.20 1.25 1.30 1.35 i

Measured to Predicted CHF -

r l

l C-6 l

~

l

FIGURE C-3 Measured to Predicted CHF Versus Mass Velocity 2.0-Mark-sw cata Base

~

M t

O s 1.5-u r

O d

8

_ ' U h_

an rg, o

an O

g 0

o h

g O

S O

i C

O l

e

- 0.5-C H

f 0.0-,

0 1

2 3

{

Uoss Yelocity, Ulba/ht-il2 1'

FIGURE C-4

~

Measured to Predicted CHF Versus Pressure Mark-BW Data Base i

2.0-M e

a s 1.5-u T

O O

O n

1 y "I

}t 8

i O

C l

e d 0.5-C H

f 0.0,

0 750 1500 2250-3000 Pressure, psia

FIGURE C-5 Measured to Predicted CHF Versus Quality Mark-BW Data Base W

e 0

s 1.5-u r

d O

U U

O n tb n

o rfn a n

U U

o o

U cp a

g a

a O@

00 6

i 0

c t

e d 0.5-C H

f 0.0,

-15.0 7.5 30.0 52.5 75.0 Thermodynamic Quality at Clif, 2

\\

TABLE C-1 VIPRE-01 BWU-Z Correlation Verification CHF Test Database Analysis Results VTPRP-01 Atarinrical Romulen Number Of Data Points 530 Average M/P 1.00850 Standard Deviation 0.09217 Upper D Prime 3469.0 Lower D Prime 3407.0 D Prime Value 3453.68 Accept Normality at 5% Level Paramptor Rancen Pressure, psia 400 to 2465 Mass Velocity, Mlbm/hr-ft' O.36 to 3.55 Thermodynamic Quality at CHF less than 0.74 Thermal-Hydraulic Computer Code VIPRE-01 Spacer Grid Mark-BW 17x17 Design Limit DNBR, VIPRE-01 1.18 C-10

I TABLE C-2 McGuire/ Catawba SCD Statepoints Core Inlet Stpt Power

• RCS Flow Pressure Temperature Axial Peak Radial Peak h

(% RTP)

(K com)

(osia)

('F )

(F 0 Z)

(FAH) 1 2

3 4

5 6

7 8

10 i

11 12 13 t

14 15 16 l

17 1

18 19 20 21 22 23 24 i

100% RTP = 3411 Megawatts Thermal C-11

TABLE C-3 McGuire/ Catawba Statistically Treated Uncertainties Standard Type of Parameter Uncertainty / Deviation Distribution Core Power

+/- 2% / +/- 1.22%

Normal Core Flow Measurement

+/- 2.2% / +/- 1.34%

Normal Bypass Flow

+/- 1.5%

Uniform Pressure

+/- 30 psi Uniform Temperature

+/- 4 deg F Uniform FNAH Measurement

+/- 4.0% / 2.43%

Normal FE

+/- 3.0% / 1.82%

Normal AH Spacing

+/- 2.0% / 1.22%

Normal

+/- 4.41% / 2.68%

Normal F3 Z

+/- 6 inches Uniform DNBR Correlation

+/- 15.25% / 9.27%

Normal Code /Model

[

]

Normal Percentage of 100% RTP (68.22 MWth wherever applied).

c-12

TABLE C-3 Continued McGuire/ Catawba Statistically Treated Uncertainties Parameter Justification Core Power The core power uncertainty was calculated by statistically combining the uncertainties of the process indication and control channels.

The uncertainty is calculated from normally distributed random error terms such as sensor calibration accuracy, rack drift, sensor drift, etc. combined by the square root sum of squares method (SRSS).

Since the uncertainty is calculated from normally I

distributed values, the parameter distribution is also normal.

Core Flow Measurement Same approach as Core Power.

l Bypass Flow The core bypass flow is the parallel core flow paths in the reactor vessel (guide thimble cooling flow, head cooling flow, fuel assembly / baffle gap leakage, and hot leg outlet nozzle gap leakage) and is dependent on the driving pressure drop.

I Parameterizations of the key factors that control l

AP, dimensions, loss coefficient correlations, and the effect of the uncertainty in the driving AP on the flow rate in each flow path, was performed.

The dimensional tolerance changes were combined with the SRSS method and the loss coefficient and driving AP uncertainties were conservatively added to obtain the combined uncertainty.

This uncertainty was conservatively applied with a uniform distribution.

l l

Pressure The pressure uncertainty was calculated by statistically combining the uncertainties of the process indication and control channels.

The uncertainty is calculated from random error terms such as sensor calibration accuracy, rack drift, sensor drift, etc. combined by the square root sum l

of squares method.

The uncertainty distribution was conservatively applied as uniform.

l l

Temperature Same approach as Pressure.

c-13 l

l l

l

l TABLE C-3 Continued McGuire/ Catawba Statistically Treated Uncertainties 1

Parameter Justification N

F AH Measurement This uncertainty is the measurement uncertainty for.

the movable incore instruments.

A measurement uncertainty can arise from instrumentation drift or reproducibility error, integration and location error, error associated with the burnup history of the core, and the error associated with'the l

conversion of instrument readings to rod power.

The uncertainty distribution is normal.

FE This uncertainty accounts for the manufacturing AH variations in the variables affecting the heat L

generation rate along the flow channel.

This conservatively accounts for possible variations in the pellet diameter, density, and U235 enrichment.

This uncertainty distribution is normal and was conservatively applied as one-sided in the analysis l

to ensure the MDNBR channel location was consistent for all cases.

Spacing This uncertainty accounts for the effect on peaking of reduced hot channel flow area and spacing between assemblies.

The power peaking gradient becomes steeper across the assembly due to reduced flow area and spacing.

This uncertainty l

distribution is normal and was conservatively l

applied as one-sided to ensure consistent MDNBR I

channel location.

I

(

Fz This uncertainty accounts for the axial peak l

prediction uncertainty of the physics codes.

The uncertainty distribution is applied as normal.

l Z

This uncertainty accounts for the possible error in interpolating on axial peak location in the maneuvering analysis.

The uncertainty is one half of the physics code's axial node. The uncertainty I

distribution is conservatively applied as uniform, 8

l C-14

1 l

l I

4 TABLE C-3 Continued McGuire/ Catawba Statistically Treated Uncertainties Parameter Justification DNBR Correlation This uncertainty accounts for the CHF correlation's ability to predict DNB.

The uncertainty distribution is applied as normal, Code /Model This uncertainty accounts for the thermal-hydraulic code uncertainties and offsetting conservatisms.

This uncertainty also accounts for the small DNB prediction differences between the various model sizes.

The uncertainty distribution is applied as normal.

l i

I C-15

TABLE C-4 McGuire/ Catawba Statepoint Statistical Results BWU-Z Critical Heat Flux Correlation 500 Case Runs Coefficient Statistical of Variation DNBR Statenoint #

Mean g

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 C-16

4 TABLE C-4 Continued McGuire/ Catawba Statepoint Statistical Results BWU-Z Critical Heat Flux Correlation 5000 Case Runs Coefficient Statenoint #

Mean g

of Variation DNBR 1

7 9

12 l

c-17 l

1 TABLE C-5 McGuire/ Catawba Key Parameter Ranges Parameter Maximum Minimum Core Power (% RTP)

Pressure (psia)

T inlet (deg. F)

RCS Flow (Thousand GPM)

FAH, Fz, Z All values listed in this table are based on the currently analyzed Statepoints.

Ranges are subject to change based on future statopoint conditions.

l l

l C-18 l

REFERENCES C-1.

PPC-NE-2004P-A, McGuire and Catawba Nuclear Stations Core Thermal-j Hydraulic Methodology Using VIPRE-01, December 1991.

C-2.

The BWU Critical Heat Flux Correlations, BAW-10199-P, Babcock and l

Wilcox, Lynchburg, Virginia, December 1994 (SER received April 5, 1996).

C-3.

VIPRE-01: A Thermal-Hydraulic Code For Reactor Cores, EPRI NP-2511-CCM-A, Vol. 1-4, Battelle Pacific Northwest Laboratories, August 1989.

l l

l C-19