ML20081M285
| ML20081M285 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/21/1983 |
| From: | SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML17334A488 | List: |
| References | |
| XN-NF-621(NP)(A)-R01, XN-NF-621(NP)(A)-R1, XN-NF-621-(NP)(, NUDOCS 8311170200 | |
| Download: ML20081M285 (113) | |
Text
{{#Wiki_filter:- - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - XN NF 621 (NP) (A) [ REVISION 1 f I 1 EXXON NUCLEAR DNB CORRELATION I FOR PWR FUEL DESIGNS i l l OCTOBER 1983 I I W I l ERON NUCLEAR COMPANY,Inc. 1 l "8A'il88Rolaba8;s P PDR
[ 4-XN-NF-621(NP)(A) [ Revision 1 Issue Date: 10/21/83 [ [ [ EXXON NUCLEAR DNB CORRELATION FOR PWR FUEL DESIGNS { ( ( This is the approved version of Document XN-NF-621(NP)(A), Revision 1, and has been prepared in accordance with NRC guidance. ( { l ERON NUCLEAR COMPANY,Inc. (
i
- puo
. ' '/
w#'o, UNITED STATES
- l' '
g NUCLEAR REGULATORY COMMISSION g . . . j WASHINGTON, D. C. 20555
% .' .v
/ APR 12 m
- o
( Dr. Richard B. Stout, Manager Exxon Nuclear Company 2101 Horn Rapids Road P. O. Box 130 Richland, Washington 99352
Dear Dr. Stout:
Subject:
Acceptance for Referencing of Licensing Topical Report XN-NF-621(P), Revision 1, " Exxon Nuclear DNB Correlation for PWR Fuel Designs" ( We have completed our review of the subject topical report submitted May 5,1982 by Exxon Nuclear Company (ENC) letter GF0:034:82. We find this report is acceptable for referencing in license applications for LWR Plants 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. y We do not intend to repeat our review of the matters described in the report and found acceptable when the report appears as a reference in license applications except to assure that the material presented is applicable to the specific plant involved. Our acceptance applies only to the matters described in the report. In accordance with established procedures (NUREG-0390), it is requested that ENC publish accepted versions of this report, proprietary and non-( proprietary, within three months of receipt of this letter. The accepted versions should 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 . [ ( [ ( f -
Dr. Richard B. Stout APR 12 B83 r (. Should our criteria or regulations change such that our conclusions as to the acceptability of the report are invalidated, ENC and/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,
\
0-Cecil 0. Thomas, Chief Standardization & Special Projects Branch Division of Licensing
Enclosure:
As stated
?;.
.. s
)
{
w C L: 1 INTRODUCTION In XN-NF-621, Revision 1,' Exxon Nuclear Company (ENC) presented the XNB critical heat flux (CHF) correlation which will be used to assess the thermal margin of ( pressurized water reactors (PWRs). The XNB is an empirical relationship which specifies CHF (i.e., the heat flux at which departure from nucleate boiling, DNB, occurs) as a function of local coolant conditions and fuel assembly { geometry. It is based on 14 test series with a total of 714 data points and three different PWR fuel vendor designs. The 14 test series include variations in grid design, heated length, grid span, rod diameter, and axial and radial power distributions. The local coolant conditions in the rod bundle were calculated using the XCOBRA-IIIC computer code which is described in XN-NF-75-21(P) and the range of coolant conditions tested were typical of an operating PWR. [. . '
\' Based on the XNB's ability to predict the test data, Exxon has proposed a departure from nucleate boiling ratio (DNBR) limit of 1.17 for the correlation.
This limit corresponds to a 95% probability of not experiencing DNB at a 95% confidence level. The comparable value for the W-3 correlation, which is ( presently used by ENC, is 1.30. S ( [ [ [ (~
~
[ Exxon SER 1-1 I _- - - - - - - - - -
L ( . L , { L 2 DESCRIPTION OF' CORRELATION The basic form of the XNB correlation is 17 follows: 4" uncorrected = A + B HLOC eq. (1) where A = f (pressure, mass velocity, inlet subcooling) B = f (pressure, mass velocity, local enthalpy) HLOC = Reduced local enthalpy
= Local Enthalpy/906.00 All'of the parameters used in the XNB are reduced using the critical properties
( of water (i.e., the water properties at the critical pressure, 3208.2 psi) [ and using the above method for HLOC. Additional factors are used as part of the correlation to account for non-uniform axial power distributions, geometry differences such as spacer pitch and mixing vane loss coefficients, and differences in heated lengths. The final form of the XNB is:
,. 5 - k. ,
q" critical = (q" uncorrected) Comcdon Factors eq. (2) ( The procedure for using the XNB is to initially calculate the heat flux using equation (1), determine the appropriate correction factors, calculate CHF [ using equation (2), and determine the DNBR, which is the ratio of the actual heat flux to predicted CHF. The ranges over which Exxon is requesting the XNB be applied (Chandler; January 6, 1983) are: Pressure (psia) 1395 - 2425 2 ( Local Mass Velocity (M1bm/hr-ft ) 0.92 -3.04
. Local Enthalpy (Btu /lb) 594.85 - 821.24 Local Quality +0.3
{ Heated Length (inches) 66 - 168 Grid Spacing (inches) 14.3 - 22.0 ( - Inlet Subcooling (Btu /lb) 37.2 - 336.34 Exxca SER 2-1 I
I \ . It will also be used for the following geometries: (" k Vendors: Exxon Nuclear Combustion Engineering Westinghouse Fuel Design: Non-Mixing Vane Mixing Vane Equivalent Hydraulic 0.177 - 0.612 Diameter (inches) Equivalent Heated 0.463 - 0.528 Diameter (inches) The test series and their associated fuel rod arrays are: Vendor Rc,d Array Test Series Westinghouse / 14x14, 15x15 ENC-3, 4, and 5 Exxon ROSAL-2, 4, 7, and 8 Exxon 17x17 ENC-6 Cumbustion 16x16 CE-47, CE-59 Engineering I Westinghouse 17x17 WH-162 and 164 I I i I l
~
l
,r -
i
\
Exxon SER 2-2
f ~ L (( 3 STAFF EVALUATION 3.1 Scope of Review ( The st ff review of XN-NF-621, Revision 1 included an independent-audit of the subchannel calculations performed to determine the local coolant conditions in
. the rod bundle for all 714 data points. This was performed using the COBRA-IV computer code which was derived from and is an ancillary of the COBRA-IIIC prog am. Our review also included a statistical analysis of the calculated -
results and a review of the methodology used in combining the XCOBRA-IIIC code and the correlation. During the review, requests were made for data clarifica-( tion and additional or corrected information was received in several areas. The above reviews were performed by the Idaho National Engineering Laboratory { (INEL) under the direction of a cognizant staff member. f.
'** 3.2 Results of Audit Calculations
( The results of the INEL audit calculations are presented in Tables 1 and 2. Table 1 is a comparison of the local conditions at which CHF was predicted as determined by the XCOBRA-IIIC and COBRA-IV codes for a limited number of data ( points. The comparison indicates good agreement between the two codes and either could be used to establish the local conditions required for the develop-ment of a CHF correlation. ( Table 2 is a comparison of the mean and standard deviation for each of the data sets and the total population. This comparison shows good agreement for f ^the overall value; but contains discrepancies in many of the individual data sett. The possible ramifications associated with these differences are de-scribed in the statistical analysis discussion contained in this report, [ f . Exxon SER 3-1
During our review, the staff requested that Exxon provide a description of how the local conditions for the XNB were determined including a discussion 4 of the subchannel code used, subchannel modeling, axial nodalization, and input assumptions. Exxon responded that the XCOBRA-IIIC code was used to calculate the local coolant conditions. XCOBRA-IIIC is a derivative of the COBRA-IIIC code which was developed at Battelle Pacific Northwest Laboratory. The modifications made by Exxon to COBRA-IIIC include minor improvements in the solution technique, the addition of calculational options, and operational modifications such as streamlining code input. Exxon further stated that the friction factors used were determined from pressure drop measurements performed on ENC test sections or estimated for geometries for which ENC does not have detailed test data. These loss coefficient estimates are based on the experience gained from measuring actual fuel bundles of Westinghpuse or Combustion Engineering (C-E) designs. They also reported that sensitivity studies of CHF test data showed negli-gible influence on predicted conditions when the form loss coefficients were varied by as much as 15L (, The mixing values (h) chosen were based on spacer design and are dependent g on a particular fuel type. These values were determined experimentally for the i ENC designed fuel while for non-Exxon fuel a lower bounding value was used for mixing vane grids. For example, in analyzi ig, the Westinghouse "L" grid design a lower value of 0.010, which was obtained from WCAP-8030-A, was used. Based on our review of the above information, the staff concludes that the approach taken by Exxon in determining the local conditions used in developing the XNB correlation are acceptable. The XCOBRA-IIIC code is still under staff review, and any limitations resulting from this review will be addressed in our safety evaluation report on XN-NF-75-21(P), Revision 2. l The INEL audit calculations were performed using the same friction factor correlation, two phase flow correlation, rossflow resistance, momentum turbulent mixing factor, pitch to length parameter, inlet enthalpy and s
. inlet mass velacity as Exxon.
Exxon SER 3-2
f I' . L Our review also included an analysis of the correction factors used in the XNB development and the determination of these factors in actual reactor application.
. Based on this review, we have concluded that the method used to calculate these parameters and their values used in determining the DNBR limit are acceptable.
However, it is the opinion of the staff and our consultant that a change in these parameters, such as determining their values using a prototype and then [ a full scale bundle, may increase the uncertainty in both the code's prediction of local coolant conditions and the correlations prediction of CHF. This may significantly alter the statistical analyses on which the DNBR limit is based. Therefore, we conclude that t.he values of these parameters used in the develop-ment of the XNB must be used in licensing analyses. ( For the uniform heat flux tests, ENC used the erd of the heated length as the CHF location while the experiments showed that for the sasbe tests, CHF occurred upstream of the end of the heated length. When asked to justify using this technique in determining the DNBR Exxon' responded that the worst local condi-
; tions calculated for a bundle having a uniform axial power distribution (APD) are at the end of the heated length. In order to maintain a consistent path between test analysis and reactor design and based on the fact that the DNBR
( location in a reactor is determined by the code and is not known apriori, the procedures used to determine the DNBR for those tests where burnout occurred upstream of the heated length is acceptable. We have reviewed the additional ( information provided by ENC and have concluded that the method used by Exxon in determining DNER is acceptable since the DNBR limit is dependent on the ability of the subchannel code to predict local conditions which produce CHF. ( An additional area of concern raised by the staff on the uniform heat flux
; tests was why CHF occurred at the thermocouple upstream of the end of the f heated length rather than at the end of the heated length where the highest quality region should occur. Exxon stated that burnout is a function of the
[ [(_ . . Exxon SER 3-3
l-location of the spacer grid and that the grids will improve heat transfer for (~ a distance of 20 or more rod diameters downstream of the spacer. Because the Y- spacer was located slightly downstream of the end of the heated length, heat
]
transfer above the spacer would improve while the local hydraulic conditions downstream of the grid would be more severe. Therefore, for the experimental data in question, the effects of the spacer grid dominated the occurrence of CHF evqn though a higher quality may occur at the end of the test bundle. The staff has reviewed this information and concludes that ENC has acceptably addressed our concerns on this issue. Finally in the area of test procedures, the staff requested that Exxon provide a discussion on how the rate of power was increased, what post-test inspections were performed, and what, if any, duplicate runs were made to establish continued integrity of the test bundle. In response to this concern, ENC stated that the power was manually raised in the CHF tests by an increment of less than 1% and held constant until conditions became stable. This process was repeated until CHF occurred. They further stated that duplicate runs were made to establish continued integrity. As an example, they cited the ENC-6 tests, f where replicate points were taken during the test and one in between point was b taken at the end of the test to confirm continuity and consistency of the test data from beginning to end. At the end of the tests, post-test inspections were performed and, for example, on the ENC-6 bundle there were no visible signs of hot spots on the rods. Based on our review of this information, the staff has concluded that the CHF tests were performed in an acceptable manner. Our review of the statistical characterization of the XNB results dealt mainly with the method used by Exxon to statistically analyze the data and a review of the analyses. The statistical method used by ENC was to evaluate the predicted-to-measured (P/M) ratio of CHF data. Since in previously approved correlations, the measured-to predicted (M/P) ratio was used to determine the 95/95 limit, Exxon was asked to justify their technique. ENC responded that the procedure used in determining the 95/95 limit assumed a normal distribution. Transforming the data from P/M to M/P yields two distribu-tions for comparison, both of which may be normal or both may depart from
, normality. As a verification on the 95/95 limit for the P/M data, Exxon ]
Exxon SER 3-4
{ ' / performed a distribution free estimate of the limit and determined the value L - to be 1.177. For the reverse ratio, and using their original statistical approach, Exxon calculated that 95/95 limit for the M/P data, when a normal ( distribution is assumed, is 1.191. ENC further stated that the non parametric estimate of the 95/95 limit, 1.177, { does ngt make complete use of the actual distribution, and therefore this limit will bound the 95/95 limit obtained from the actual distribution. By considering the first four moments of the P/M data ENC found that the actual distribution is a gamma distribution. On the other hand, the use of the M/P f data is overly conservative since, the actual value of the 95/95 limit for the P/M data, when the appropriate distribution is used, lies at some value below the non parametric limit of 1.177. ENC also stated that the DNBR reported for ( licensing analyses is defined as P/M ratio. Based on our review of the above information, the staff has concluded that the analysis of the P/M data is acceptable. As part of the review, the staff requested that Exxon demonstrate that each of the samples, e.g., test series, belong to a single population. ENC responded ( by initially performing a Bartlett test for homogeneity of variance (Chandler; August 26, 1982). The breakdown was based on both vendor design and fuel assembly geometries. The results of this test showed that the variances do differ among geometry types. Exxon also performed a K-sample Squared Ranks test of variance using the above groupings (Chandler; August 26, 1982). Results for the population of 6 samples ( and 5 degrees of freedom indicated that at least two of the variances were unequal. By removing the ROSAL, ENC-1, and 2 data, Exxon found that there exists a significance level between 2.5% and 5.0% that the remaining data were from the same population. Finally, ENC removed the ENC-3, 4, and 5 data and analyzed the remaining population. Based on the results of the third analysis, Exxon concluded that the data comprised of 3 samples and 2 degrees of freedom were likely identical. Exxon SE, 3-5 ( (
l An analysis of the means and a comparison of variance analysis showed that for
'~ cn equivalent sample size of 83.7 with 378.7 degrees of freedom the mean is di 0.98502 with a standard deviation of 0.09847. Based on this mean and standard deviation the 95/95 DNBR limit would be 1.168.
The final analysis performed by ENC was the determination of a DNBR limit cxcludi,ng that data which had the greatest possibility of being from a different population. For all sections less the ROSAL and ENC 1 thru 5 data the DNBR limit was 1.169 while for all sections less the ENC-6, WH-162, W-164, CE-47, _ and 49 data, the DNBR limit was 1.176. ' The results of the above tests lead ENC to conclude that the data could be treated as a single population and that the 1.17 DNBR limit would cover any g deviation within the data sets. In order to ascertain the validity of these conclusions, INEL performed a series of F-tests to identify any systematic variation among the test series. The tests were performed at a 99% confidence level. Based on the F-test, INEL C,~ concluded that there was a variance among tests of different geometries. N Additionally, INEL performed a one-way analysis of variance using the ungrouped test series. For the one-way analysis, INEL used the groupings reported by ENC and calculated a F-ratio of ?4.03 for six samples with five and 708 degrees of freedom for the numerator and denominator. This result shows that there is a variance among the tests when they are grouped by geometry type. Removing data sets ( WH-162, WH-164, ENC-3, 4, and 5 resulted in an F-ratio of 2.40 with three and 392 degrees of freedom for the numerator and denominator. This indicates that the remaining data have a probability of between 5% and 10% of being in the same population. A second one-way analysis of variance was performed on the ungrouped data. The results of this test are presented in Table 3 and indicate that ENC-1, ENC 2, ENC-6, ROSAL-2, ROSAL-7, ROSAL-8, WH-162, CE-47 and CE-49 are probably of t5e same population while test series ENC-3, ENC-4, ROSAL-4, and WH-164 are Exxon SER 3-6
s d 9 J of a second population. ENC-5 is a unique test series and does not fall into
~
L either population, Using the above populations, a DNBR limit of 1.21 for the ENC-1, ENC-2, etc. population was determined while the ENC-3, ENC-4, etc. h population has a 95/95 limit of 1.133. Figure 1 is a histogram of the total data set and it shows that the overall ( populat. ion is approximately normally distributed. Histograms for the individual samples (EGG-NTAP-6167) show that ENC-3, ENC-4, ENC-5, ROSAL-4 and WH-164 are skewed to the left of the population mean, f Further analyses were performed to determine if there was a reason for the groupings obtained from the one-way analysis of variance. A number of groupings ( were examined using different bases such as rod diameter, grid spacing, radial power distribution, axial power distribution, KLOSS, and an unheated guide tube in the bundle. These studies showed no uniqueness in either grouping. A second evaluation revealed that the modeling of the guide tube was an influence in determining the above grouping. For those bundles containing an unheated ( , guide tube, CHF experimentally occurred in a channel that contained the guide
**# tube; however, in predicting CHF, Exxon often reported burnout in a channel
( other than the one with the guide tube. Since the guide tube is an unheated wall, CHF occurs at less severe local conditions and has a lower value. If { CHF is predicted in a typical channel, four heated rods, when it actually occurred in a guide tube channel, this would be nonconservative. The reason [ for .his is that the predicted local conditions are greater than the conditions which experimentally produced CHF; therefore, the analytical results show that you can go to a higher power than you actually achieved. Table 4 presents a summary of the test series that have one or more unheated guide tubes. For all of the series reported in Table 4 ENC predicted CHF in the COBRA hot channel rather than the experimental channels listed in the table. This indicates that the reason ENC-3, ENC-4, and ENC-5 do not belong to the population may be the difference in the channel for the predicted and measured CHF. Test series ENC-6 does not fall from the population because the
,.s
( w Exxon SER 3-7 m -___. _. _ . . _
l' difference between the COBRA-IV experimental hot channel and the guide tube channel is only 3.0% and the sample mean is closer to the expected mean of
" ,* ) 1. 0.
In addition to the above analyses, the INEL audit calculations revealed that the ENC-1, 2, 3, 4, 6, CE-59, and ROSAL-8 test series were biased with inlet pressure. For pressures less than 1800 psia the correlation predictions tend to be scattered about some value less than 1.0 while for data above 1800 psia the data is randomly scattered about 1.0. This indicated that the correlation under predicts CHF for the lower pressures but is reasonably accurate for pressures above 1800 psia. Based on this review, the staff has concluded that ] although these test series statistically belong to one of the two populations, excluding the ENC-5 population, the fact that they are biased with pressure / may preclude them from being placed in either population. Also, the staff statistically analyzed the six different geometry types reported by Exxon. Table 5 contains the results of our analysis based on a geometric - characterization. These results show that for the ENC-1 and -2 population the
^^ mean, standard deviation, and 95/95 limit are much greater than the mean, standard deviation, and 95/95 limit of the remaining populations when they are compared to the same parameters of the total population.
Based on our review of the ENC statistical analyses, our consultant's analyses,
]
and the result of the staff's statistical analyses, we requested additional information from Exxon which justified treating the 14 samples as one population.
}
In response to our concerns, Exxon provided plots of DNBR versus inlet pressure s for those test series that the staff felt were biased with pressure (Chandler; December 16, 1982). Based on their own presse e plots ENC concluded that there was no significant systematic trends with pressure. We have reviewed the.information submitted in the December 16, 1982 letter and have concluded that there is a small trend with pressure; however, the trend is random in nature and does not exhibit any systematic characteristics. Therefore, the
., staff concludes that the ENC-1, 2, 3, 4, 6, CE-59, and ROSAL-8 test series Exxon SER 3-8 H
s f c' [ , need not be treated as a single population due to the trends in pressure,
.{ since these trends are not systematic.
With respect to the statistical analyses, Exxon requested that the data be reviewed as two separate populations (Chandler; December 22, 1982). One of the populations would be comprised of the test series representing 16x16 and 17x17 arrays (CE-47, CE-59, WH-164, WH-162, and ENC-6) while the second popula-tion would represent the 15x15 bundles. As justification for requesting this ( breakup, ENC provided the range of test conditions and axial power distri-butions found in each population. ( A review of the 16x16 and 17x17 data base showed that only a chopped cosine and uniform axial power distribution (APD) were present. It is the position of the staff that all possible power distributions expected throughout an operating cycle be used in the development of any CHF correlation. Since the 16x16 and 17x17 do not include either an upskew or downskew APD, Exxon cannot remove those test series, e.g. the 15x15 array, that have the upskew APDs. ( Therefore, the 15x15 test series must remain in the data base until ENC pro-vides additional data for the 16x16 and 17x17 test series which contain an upskew and/or downskew APD. In a modified response (Chandler; January 3, 1983) Exxon requested that test series ENC-1 and ENC-2 be removed from the data base. The reason for elimin-( ating this data was that ENC-1 contained minimum grids that were not repre-sentative of any grid being manufactured by ENC, Westinghouse or CE while ENC-2 had a uniform axial and radial power distribution that was atypical of { , actual reactor conditions. ENC further stated that a statistical analysis of the data was performed using the populations reported by INEL. The results of these evaluations showed that the worst 95/95 limit was 1.17 for the population containing the CE-47, -59, WH-162, ENC-2, ROSAL-2, -7, and -8 test series. Based on these results, we have concluded that the proposed grouping of data which results in a DNBR limit value of 1.17 is acceptable. ( [d. [ Exxon SER 3-9
s ? I ) l L ( l 4 CONCLUSION The staff has reviewed XN-NF-621, Revision 1 and the additional supporting { information submitted by Exxon Nuclear Company. Based on this review, we have concluded that XNB correlation is acceptable for use in reactor licensing applications. We have also concluded that the 95/95 DNBR limit of 1.17 reported by Exxon is acceptable. These conclusions are based on the following: (1) The subchannel code used, XCOBRA-IIIC, is acceptable for predicting local coolant conditions used in the development of a CHF correlation. { This is based on a comparison of XCOBRA-IIIC with the staff's audit code COBRA-IV. Since the XCOBRA-IIIC is still under staff review, any limita-(' tions resulting from its u'se will be addressed in our safety evaluation report on the code. (2) An independent audit, performed by our consultant INEL, using a different
.. subchannel code yielded similar results.
(3) The DBNR data has been statistically characterized in an acceptable { i manner. (4) The 95/95 limit is based on three separate populations that were recom-mended by our consultant; therefore, the 95/95 limit of one population will be conservative when compared to the limit of a population containing all of the test data. We will require that the correction' factors used in analyzing the CHF test data and the mixing factors used in the data reduction be used in reactor design applications, since a change in these factors may alter the code and correlation uncertainties associated with the prediction of CHF. This in k turn may raise or lower the 95/95 DNBR limit. Therefore, if any of these parameters are changed, ENC must provide a description of the change and ( Exxon sER 4-1 t _ -_- --
i sufficient justification which warrants making this change. Additionally. Exxon should provide the test data which justifies using the XNB on fuel 'l designs not contained in the data base or acceptable justification on why the XN8 is applicable to this fuel type. For example, Exxon manufactured fuel for CE reactors is not present in the data base. ENC must provide additional test data for these fuel bundles or a quantified justification of the XNB's appli-cability to this bundle type. Finally, it should be noted that the DNBR limit does not include any adjustment which is required when a mixed core, e.g. a core with geometrically different fuel types, is analyzed. f J (t xxon E R 2
1 ( : L '. 6 5 REGULATORY POSITION ( , The staff concludes that the XNB CHF correlation as described in XN-NF-621, Revision 1 is accepteble for use in licensing application when it is used with the XCOBRA-IIIC code and within the range of application reported in . Section 2.2 of this safety evaluation report. We also conclude that the 55/95 limit of 1.17 associated with the XNB is acceptable. Use of the ( correlation should be within the limitations described in the previous section. ( Based on our review, the staff finds XN-NF-621, Revision 1 an acceptable and referential report with the restrictions noted in the above paragraph. ( ( (.- ( ( l [ . L L,\b Exxon SER 5-1 r -- - - - - - -
- r _
-_ - m- -- w N. ,
- 4. . , , ,
Table 1: Comparison of Local Conditions Enthalpy Quality Void Fraction Mass Flux Case XC08HA-III COBRA-IV XCOBRA-III COBRA-IV XC08RA-IIIC COBRA-IV XCOBRA-IIIC C08RA-IV l 1.9434 ENC-3-63 656.57 656.49 0.077 0.077 0.'610 0.594 1.9046 ENC-4-28 703.28 705.78 0.167 0.167 0.709 0.712 1.4897 1.5210 I ENC-6-42 616.44 628.00 0.00 0.007 0.318 0.350 2.8655 2.8998 ROSAL-2-18 612.57 627.50 0.001 0.027 0.550 0.554 1.8674 1.8809 ROSAL-2-9 622.44 636.52 0.018 0.043 0.561 0.566 1.9409 1.9601 e G
I Table 2: Comparison of Mean anc Standard Deviation Test Number of Mean (Meas./Pred) Standard Deviation Section Data Points XCOBRA-IIIC COBRA-IV XCOBRA-III COBRA-IV CE-47 96 1.028 1.0300 0.0741 0.0804 CE-59 89 1.023 1.0500 0.0820 0.1020 WH-164. 53 0.950 0.9727 0.0677 0.0682 WH-162 53 0.992 1.0032 0.0845 0.0736 ROSAL-2 28 0.976 0.9995 0.118 0.0990 ROSAL-4 26 0.933 0.9689 0.0843 0.0832 ROSAL-7 11 0.970 1.0383 0.1043 0.1210 ROSAL-8 32 1.001 1.0586 0.0987 0.1070 ENC-1 28 1.040 1.0504 0.1212 0.1220 ENC-2 24 0.993 1.0119 0.1093 0.1090 ENC-3 73 0.994 0.9458 0.1029 0.0923 ENC-4 80 0.985 0.9712 0.1196 0.112 ENC-5 59 0.911 0.8956 0.0848 0.0811 ENC-6 62 0.995 1.0071 0.0749 0.0868 Total Population 714 0.985 0.99614 0.09847 0.1030 o v. 2 {
/ I - s - Table 3: One Way Analysis of Variance Probability of Being Test Series Grouping F-Ratio in Same Population ENC-1, -2, -6 ROSAL-2, -7 -8 WH-162CE-47 -59 , 2.47 1 - 2.5% ( ENC-1, -2, -4, -6 ROSAL -2, -4, -7, -8 ( WH-162, -164 CE-47, CE-59 5.57 --- ENC-1, -2, -3, -4, -6 ROSAL-2, -4, -7, -8 ( WH-162, -164 - C~.-47, -59 7.84 --- ( ENC-3, -4, -5 [,,3 ' ROSAL-4, WN-164 7.39 --- [ ENC-3, -4 ROSAL-4, WH-164 1.23 >10% ( [ . [ Y 3
I
~
i I l Table 4: Comparison of Test Series With Unheated Guide Tubes v Number of Experimental CHF Predictions COBRA-IV Channel Test COBRA-IV Other Than Series Hot Channell Hot Channel Explanation WH-162' All As expected. ENC-6 26 42 The 42 channels are 3% cooler than the hot channel. ENC-3 18 53 Five of the indications occur in a channel with 5% less power, 21 in a channel with 0.4% less power and the remaining in a channel with 23% i less power. ENC-4 30 50 Seven of the 50 indications were in a channel with 0.20% less power , while the remaining 43 wers in a channel with 22% less power. ENC-5 . 4 53 Twenty-five of the 53 indications [ .- occur in a channel with 0.9% less power while the remaining 28 are in a channel with 22% less power. CE-47 82 14 The 14 indications occur in a channel with 0.3% less power. CE-59 85 4 The 4 indications occur in a channel with 0.1% less power. 1 ENC predicts all CHFs in this channel. i {
L (. . r Table 5: Comparison of 95/95 Limit Based on Geometry ($' ' Geometry Grouping Mean Standard Deviation 95/95 Limit ( CE-47, CE-59 1.0256 0.0778 1.169 WH-162, WH-164 0.9710 0.0791 1.123 ENC-6 O.995 0.0749 1.146 ROSAL-2, 4, 7, 8 0.9720 0.1021 1.169 ENC-1, ENC-2 1.0183 0.1173 1.259 [ L ENC-3 ENC-4, ENC-5 0.9503 0.0865 1.109 Total Population 0.985 0.0985 1.163 ( . ( ( 6, [ ( ( ( ( . 5 V r (k' [ 5 l
i. 9 [ 6 REFERENCES 6.1 Topical Reports [ XN-NF-621, Revision 1 " Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, April 1982. XN-NF-75-21(P), Revision 2, "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operration," Exxon Nuclear Company, September 1982. k WCAP-8030-A, " Application of Modified Spacer Factor to L Grid Typical and Cold Cell DNB," Westinghouse Electric Corporation, January 1975. 6.2 Other References t { l' ' Ambrosek, R. G. , C-C Tsai, and R. D. Wadkins, " Exxon Nuclear DNB Correlation Review," EGG-NTAP-6167, EG&G Idaho, INC., February 1983. J.C. Chandler (ENC) to H. Bernard (NRC),
Subject:
"XN-NF-621(P), ' Exxon
( ' Nuclear DNB Correlation for PWR Fuel Designs,' Revision 1," August 26, 1982. J.C. Chandler (ENC) to J.J. Holonich (NRC),
Subject:
"XN-NF-621, Revision 1, ' Exxon Nuclear Company DNB Correlation for PWR Fuel Designs,'
April 1982," December 9,1982.
?
r l J.C. Chandler (ENC) to L.E. Phillips (NRC),
Subject:
"XN-NF-621, Revision 1,
' Exxon Nuclear Company DNB Correlation for PWR Fuel Designs,' April 1982,"
December 16, 1982. ( ( . (/ l ' Exxon SER < 6-1
i J.C. Chandler (ENC) to L.E. Phillips (NRC),
Subject:
"XN-NF-621, Revision 1, T ' Exxon Nuclear Company DNB Correlation for PWR Fuel Designs,' April 1982 " - [:: December 23, 1982.
J.C. Chandler (ENC) to L.E. Phillips (NRC),
Subject:
"XN-NF-621, Revision 1,
' Exxon Nuclear Company DNB Correlation for PWR Fuel Designs,' April 1982,"
January.3, 1983. J.C. Chandler (ENC) to L.E. Phillips (NRC), Subject.: "XN-NF-621(P), Revision 1, ' Exxon Nuclear DNB Correlaticn for PWR Fuel Designs,' April 1982," January 6, 1983. 4 E!;..- 1 ((.' Exxon SER 6-2
s f L
?N-NF-621 (NP)( A) nevision 1
( EXXON NUCLEAR DNB CORRELATION FOR PWR FUEL DESIGNS ( Prepared by: Mu 7 Apak 81 ( , R. B. MacdW T Date Approved by: M I. W. Patten, Manager, Fuel Testing
///A/fD Date t #$a(W J. F. 'Patterson, Manager da/n Date F 1 Development and Testing
- h-J. N. prgan, ManageM fuel Licensing VAz/r>
Date S N$E l-G. A. f ,Mpager Date - Fuel ineerimg & Technical Services
}af.Pk G. L. Ritter, Manager
% /r v Date Process and Engineering Development
[ jos k
l l
' NUCLEAR REGULATORY COMMIS$lON DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was sterived through research and development progroms sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reloart fust or other techncal services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information conteined herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are cussomers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any persoa acting on its behalf: A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infritage privately owned rights;
- or B. Assurnes any liabilities with respect to the use of, or lui darrages resulting from the use of, any information, an-paratus, method, or process disclosed in this document.
XN NF F00, 766
s I L l r L -i- XN-NF-621 (NP)( A) Revision 1 k EXXON NUCLEAR DNB CORRELATION { FOR PWR FUEL DESIGNS TABLE OF CONTENTS { Page
1.0 INTRODUCTION
AND
SUMMARY
....................................... 1-1
(. 2.0 ENC DEPARTURE FROM NUCLEATE BOILING CORRELATION ................ 2-1 2.1 XNB CORRELATION .......................................... 2-1 2.2 NON-UNIFORM AXIAL HEAT FLUX FACTOR ....................... 2-2 2.3 GEOMETRIC CORRECTORS ..................................... 2-3 2.3.1 Spacer' Pitch Factor .............................. 2-3 [ 2.3.2 Mixing Vane Factor ............................... 2-3 2.3.3 Length Factor .................................... 2-4 2.4 PROCEDURE FOR USING THE XNB CORRELATION AND CORRECTION FACTORS TO PREDICT DNB HEAT FLUX ......................... 2-4 ( 3.0 0 COMPARISON OF XNB AGAINST EXPERIMENTAL DNB DATA ................ 3-1 3.1 EXXON NUCLEAR PWR DNB TEST DATA .......................... 3-1 3.1.1 Test Sections 1 & 2 .............................. 3-2 3.1.2 Test Sections 3 a 4 .............................. 3-2 3.1.3 Test Section 5 ................................... 3-3 ( 3.1.4 Test Section 6 ................................... 3-3 ( l [ _ _ _ _ _ _ _ -
XN-NF-621 (NP)(A) Revision 1 3.2 C. FIGHETTI AND D. REDDY DNB DATA ........................ 3-4 3.2.1 Combustion Engineering DNB Test E -? a . . . . . . . . . . . . . 3-4 3.2.2 Westinghouse DNB Test Data ....................... 3-5 3.3 ROSAL, ET AL TEST SECTIONS ............................... 3-5 3.4 DISCUSSION AND CONCLUSIONS OF ANALYSIS OF DNB TEST DATA .. 3-6 3.4.1 Subch annel Mix i ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-6 4.0 STATISTICAL EVALUATION ......................................... 4-1
5.0 REFERENCES
..................................................... 5-1 APPENDICES A DATA
SUMMARY
FOR TEST SECTIONS - ENC ........................... A-1 B DATA
SUMMARY
FOR TEST' SECTIONS - ROSAL ......................... B-1 C DATA
SUMMARY
FOR TEST SECTIONS - CE/WH .................... .... C-1 {
s
-111- XN-NF-621(NP)(A)
Revision 1 f L LIST OF TABLES Page ( 1.1 Summary of DN8 Data Analyzed ................................... 1-3 3.1 Sumary of Test Condt ion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-8 ( 3.2 Statistical Summary ............................................ 3-11 [ LIST OF FIGURES 1.1 Comp ar i son of Heat F lux - ALL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 [ 3.1 Test Asse: ably Geometry and Local Power Distribution for i ENC 1 & 2 Test Assemblies ...................................... 3-12 3.2 Spacer, Pressure Tap and Thermocouple Locations for ENC 1 Test
~
3-13 3.3 Spacer, Pressure Tap and Thermocouple Locations For ENC 2 Test 3-14 [ 3.4 Comparison of Heat Flux - ENC-1 ................................ 3-15 3.5 Comp arison of Heat Flux - ENC-2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-16 3.6 Test Assembly Geometry and Local Power Distribution for {- ENC 3 & 4 Test Assemblies ...................................... 3-17 3.7 Spacer, Pressure Tap and Thermocouple Locations for Tests ENC 3 & 4 ...................................................... 3-18 3.8 . Comparison of Heat Flux - ENC-3 ................................ 3-19
-3.9 Comp ar i son of Heat Fl ux - ENC-4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-20 3.10 Test Assembly Geometry and Local Power Distribution for ENC-5 Test Assembly .................................................. 3-21
( 3.11 Spacer, Pressure Tap and Thermocouple Locations for ENC-5 Test Assembly ....................................................... 3-22 3.12 Comparison of Heat Flux - ENC-5 ................................ 3-23 3.13 Test Assembly Geometry and Local Power Distribution for ENC-6 [. Test Assembly .................................................. 3-24 3.14 Axi al Power Distribution for ENC-6 Test Section . .. . . . . . . . . . . . . . 3-25 3.15 Spacer, Pressure Tap and Thermocouple Location for ENC-6 Test [ Assembly ....................................................... 3-26 3.16 Comparison of Heat Flux - ENC-6 ................................ 3-27 . 3.17 Test Assembly N metry and Local Power Distribution for CE-47 3-28 ( Test Section ................................................... 3.18 Test Assembly Geometry and Local Power Distribution for CE-59 Test Section ................................................... 3-29 3.19 Spacer and Thermocouple Locations for CE-59 Test Section and Spacer Locations for CE-47. Test Section ........................ 3-30 3.20 Axial Power Distribution for CE-59 Test Section . . .. . .. . .. . . . . .. 3-31 3.21 Comparison of Heat Flux - CE-47 ................................ 3-32 3.22 Compari son of Heat F lux - CE-59 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ( 3.23 Test Assembly Geometry and Local Power Distribution for W-162 3-33 Test Section ................................................... 3-34 L L - - - _ -
l
-iv- XN NF 621(NP)( A)
Revisioa 1
]
LIST OF FIGURES Page 3.24 Test Assenbly Geometry and Local Power Distribution for W-164 Test Section ................................................... 3-35 3.25 Spacer and Thermocouple Locations for Test Sections W-162 and W-164 .......................................................... 3-36 3.26' Axial Power Distribution for W-152 and W-164 Test Sections ..... 3-37 3.27 Comp ari son of Heat Fl ux - WH-162 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-38 3.28 Comparison of Heat Flux - WH-164 ............................... 3-39 3.29 Test Section Geometry and Local Power Distributions for ROSAL 2, 6, 7 and 8 Test Assemblies ..................................... 3-40 . 3.30 Spacer and Thermocouple Locations for ROSAL 2, 6, 7 and 8 Test - Sections ....................................................... 3-41 3.31 Axial Power Distributions for ROSAL Data ....................... 3-42 3.32 Comparison of Heat Flux - ROSAL-2 .............................. 3-43 3.33 Comparison of Heat Flux - ROSAL-4 .............................. 3-44 3.34 Comparison of Heat Flux - ROSAL-7 .............................. 3-45
.3.35 Comparison of Heat Flux - ROSAL-8 .............................. 3-46 i .
l l
L I L 1-1 XN-NF-621(NP)( A) Revision 1
1.0 INTRODUCTION
AND
SUMMARY
Exxon Nuclear presents in this report a new correlation for assessing ( thermal margin in p essurized water reactors (PWRs). The thermal margin in pressurized water reactors is assessed with a correlation of the local fluid conditions which result in a sudden rise in fuel rod temperature. This temperature rise is due to a degradation of heat transfer at the rod surface which as commonly known as d_eparture from nucleate b, oiling (DNB) or c,ritical heat flux (CHF). The correlation described in this report, the XNB correla-tion, has been compared with data gathered at Columbia University (2,3) with test assemblies I;epresenting several different designs, as summarized in Table i.i. The local fluid conditions which lead to DW have been predicted by a subchannel analysis of the test assemblies. This analysis is performed with ( the XCOBRA-IIIC kl) computer code, which performs a simultaneous solution of f equations representirq the conservation of mass, momentum, and energy. The calculated local fluid conditions were used as the correlative basis in i predicting the rod surface heat flux dich results in DNB. The XNB correla-Lion is comprised of a base correlation with a correcting term for non-uniform axial heat flux profile, cortecting terms for fuel length, spacer pitch, and f mixing vanes. {
1-2 XN-{ N{9n Rey -62 p (NP)(A)
]
for each data point in the data base, the ratio of the heat flux pre-
]
dicted by the XNB correlation to that measured in the testing (DNB heat flux ratto) has been determined. A comparison of the predicted heat flux to measured heat flux for all data is shown in Figure 1.1. The average DNS ratio as well as the standard deviation have been determined to assess the accuracy of the XNB correlation. This comparison shows that a fuel rod operating with a minimum DNB ratio (HDNBR) of 1.16 is assured that with 95% confidence, there is a 95% probability of avoiding DNB. ] t {
m_ v c-- c , , , l Table 1.1 Summary of DNB Data Analyzed Heated Gr2d Pod Test Grad
- Length Span Diameter Power Distribution DNBR Number Bundle Type ( feet) (2nch) (anch) Axial Radial- Mean Points l
CE-47 NV 12.5 14.30 .382 UNIFORM .97-1.14 1.028 0.0741 96 CE-59 NV 12.5 14.30 .382 'C050 .96-1.20 1.023 0.0820 89 kH-162 HV 14. 22.0 .374 COSU .95-1.10 0.992 0.0845 53 WH-164 HV 14. 22.0 .374 COSU .94-1.10 0.9 0 0.0677 53 ENC-6 HV 12. 20.5 .360 COSU .97-1.10 0.995 0.0749 62 ENC-1 HG 6. 15.5 .413 UNIFORM UN1f0RH 1.029 .1186 28 ENC-2 NV 6. 15.5 .413 ' UNIFORM UNIFORM .983 .1084 24 ENC-3 HV 6. 15.7 .421 UNIFORM .95-1.1 .939 .0895 73 ENC-4 HV 6. 15.7 .421 UNIFORM .95-1.1 .985 .1196 80 ENC-5 HV 5.5 26.2 .424 UNIFORM .95-1.08 .915 .0843 59 HOSAL-2 HV 8. 20. .422 USINU .95-1.15 .976 .1118 28 ROSt.L-4 NV 8. 20. .422 USINU .95-1.15 .933 .0843 26 ROSAL-7 HV 8. 20. .422 COSU .98-1.05 .970 .1043 11 [f ROSAL-8 HV 8. 26. .422 COSU .98-1.09 1.001 .0987 32 5[f
.984 0.0964 714 S f$
.g
.3
- Legend: NV = Non-Vaned g HV = H1xang Vane
- HG = Hanimum Grad i
- e. .
/ ,// '
\
,.* ////- '
,,! e /,.'/l d
/ e .- /
,> +pv+$
)
s ~ s A s \ ' % ' > > N'N 7 e -
' s o
+
,, 9 O
e
)
W 3
% 8, o ]-
/
# h
+
- 9 +
j
- e. b l g e w
x V 6, g -.- - d
.-- O e4 4 O
b*# Q*
-. e /
# 6 e
8.
- . 0 0
- 0 x
^
e
+
' ' O
. .J g
s s , - s \ W d %* 5.o *D b e ** O'
#%. sM
{
s / 2-1 XN-NF-621 ( NP)( A) L Revision 1 2.0 ENC DEPARTURE FROM NUCLEATE BOILING CORRELATION f The onset of boiling transition or Of6 la chat acterized by an abrupt decrease in the bolling heat transfer coefficient due to a change in heat transfer mechanisms. This is indicated by a temperature excursion of the heating surface. The maximusa heat flux attained before boiling transition is called the c_ritical heat flux (Chf) or d_eparture from nucleate boiling (DNB). ( For PWR operation, DNB heat flux is predicted using the XNB correlation plu:s correctors for axial heat flux distribution and geometry. 2.1 XNB CORRELATION The XNB correlation is based upon subchannel analysis of experi-mental DNB data which was used to determine the effects of local enthalpy, The analysis of experimental ( mass velocity, and pressure on DNB heat flux. data resulted in the following empirical correlation: ( QPRED = A + B
- HLOC (2.1) where:
{ ( PP l' 200.2 = REDUCED PRESSURE G , Poi i a 0C11Y/G, = REDUCED HASS VELOCITY { HSLt = (HI - HIN)/906.00 = REDUCED INLET SUBC00 LING QPRED = PREDICTED CRITICAL HEAT FLUX IN HBTU/HR FT2 HLOC = H/906.00 = REDUCED LOCAL ENTHALPY f G, = 1.0 HLB /HR FT2 (
2-2 XN-NF-621 ( NP)( A) Revision 1
)
2.2 NON-UNIFORM AXIAL HEAT FLUX FACTOR The flux shape factor F developed by Tong et al(12) provides, in part, an estimate of the effect of non-uniform ex1al in the prediction of DNB heat. flux. This factor las lerit
- C o l q"(z) [exp(-C (1 erit-Z))]dz F
loc D - exP(-C lcrit)3 where: C in' q" loc = 1 cal heat flux at Z = i crit X ,t= local quality at Z = I ctit G = mass velocity at Z = 1 crit {
)
1 [ w
s ? L r 2-3 XN-W-621 ( NP )( A) k Revision 1 ( The predicted heat flux for a non-uniform axial is ( QPREDNU = QPRED/FAXIAL (2.4) 2.3 GEOMETRIC CORRECTORS { Comparison of data among sets which differed because of bundle length, mixing vanes, or spacer pitch resulted in several multipliers modeling these effects. 2.3.1 Spacer Pitch Tactor The spacer pitch corrector was estimated as a linear fit { among data from spacers with pitches ranging from 14.25 inches to 26.2 inches. ( (2.5) where: CAP is the spacer pitch in inches, and SPC is the multiplier. Therefore, ( ( 2.6) 2.3.2 Mixing Vane Factor The mixing vane factor is based upon the spacer pressure drop. Exxon Nuclear ensures hydraulic compatibility with fuels designed by other vendcro by measuring pressure drop of full sized fuel bundles. Loss ( coefficients for spacers are then determined from the pressure drop measure-ments. The mixing vane factor ist [
2-4 XN-NF-621 (NP )( A) Revision 1 sere:
)
2.3.3 Length Factor Finally, a length correction term was observed when comparing all the data from all the bundles. This corrector is: The estimate for predicted critical heat flux corrected
- for non-uniform axial, spacer pitch, mixing vanes, and length becomes:
QPREDT = QPREDMV
- CC. ( 2.10) 2.4 PROCEDURE FOR USING TE. XNB CORRELATION AND CORRECTION FACTORS TO PREDICI DN8 HEAT FLUX The following steps are required to predict heat flux to reach boiling transition (DNB) for a subchannel in a bundle with a non-uniform or uniform axial heat distribution.
s f A
s I L f 2-5 XN-NF-621 (NP)(A) Revision 1 ( a) Calculate the local subchannel average cross section values of coolant flow, enthalpy and plessure at each axial node using XCUBRA-111C.kl) Appropriate accounting for subchannel mixing in XCOBRA-IIIC is discussed in Section 3.4.1. At each axial node calculate the pr'edicted critical heat flux using Equation (2.1). b) The F factor is calculated using Equation (2.2) and is then modified by Equation (2.3). The non-uniform axial heat flux is predicted by Equation (2.4). { c) Spacer pitch factor is calculated using Equation (2.5). The * ( predicted herst flux is calculated with Equation (2.6). d) The mixing vane factor is calculated using Equation (2.7). ( tquation (2.8) predicts heat flux. e) The factor accounting. for bundle length is calculated using { tquation (2.9) and the predicted heat flux is represented by Equation (2.10). f~ f) The DN8R is determined as the ratio of predicted heat flux to the tod heat flux. The minimum value of DNBR whether calculated for a test or reactor . operation establishes the DNB heat flux for the bundle operation condition being analyzed. For a test or experimental DNB condition, the predicted axial
- f. _ localnon of UNB determined by the preceeding approach may not always coincide with the location of the DNB detection thermocouple giving the first ON8
- l. Indication during the test. As XNB is able to predict critical heat flux curresponding to the measured critical heat flux such that the MONBR is acceptuule, the precise axial location within the test has no importance.
[ - _ - - - - -
s I. L ? [ 3-1 XN-W-621 (NP)( A) Revision 1 r L 3.0 COMPARISON OF XN8 AGAINST EXPERIMENTAL DN8 DATA Experimental DNB data are compared with predictions of DN8 using ( the XN8 correlation. The sources of DNB data include: (1) The Exxon Nuclear DNB Test Programs (2). o Minimum grid data o Non-mixing vane grid data o Mixing vane grid data { (2) C. Fighetti and D. Reddy(3) ( o Non-mixing vane grid data for Combustion Engineering design a Mixing vane grid data of Westinghouse design (3) Rosal, et al(l") o Non-mixing vane grid data o Mixing vane grid data 3.1 EXXON NUCLEAR PWR DNB TEST DATA The Exxon Nuclear DN8 test programs were conducted in the high pressure heat transfer facility at the Chemical Engineering Research Labora-tories of Columbia University. The test programs ( , ,6,0 used test assemblies of 5x5 arrays. h Characteristics of these arrays along with operating test parameter ranges are presented in Table 3.1. l l
l l 3-2 XN-NF-621 (NP)(A) Revision 1 3.1.1 Test Sections ENC-1 & 2 Two test sections, each with 25 rods of six foot length
-and uniform axial and radial profiles are included in this analysis. The characteristics of the test. sections and range of experimental conditions are shom on Table 3.1. )
The distinction between the sections was the grid design. ENC-1 used a simple support grid, referred to as a minimun grid because of the minimum impact the grid has on the flow. ENC-2 used a non-mixing vane grid
]
prototypic of production grida. The rod arrangements for the two designs are sho m as figure 3.1. Figures 3.2 and 3.3 show locations of grid spacers while a comparison of measured to predicted heat flux is illustrated on Figures 3.4 and 3.5. The mean value of DPER was 1.029 with a standard deviation of 0.1186 for ENC-1 data and 0.983 with a standard deviatlon of 0.1084 for the ENC-2 data. 3.1.2 Test Sections ENC-3 & 4
]
Two test sections, each with 21 heated rods of six foot length uniform axial and non-uniform radial profiles are included in this
~
analysis. The characteristics of the test sections and range of experimental conditions are shom on Table 3.1. The mixing vane density was the distin-guishing feature between the designs. ENC-4 used twice the number of mixing vanes as ENC-3. 1 l
s / 3-3 XN-NF-621 (NP) t A) f Revision 1 L r L The rod arrangements, grid locations, and comparison of measured to predicted heat flux are shown on Figures 3.6 through 3.9. The {. mean HDNBR was 0.939 with a standard deviation of 0.0895 for ENC-3 and was 0.985 with a standard deviation of 0.1196 for ENC-4.
~
f 3.1.3 Test Section ENC-5 This test incorporated a mixing vane spacer design with a 26-inch spacer pitch. The characteristics of the test section and range of experimental conditions are shown on Table 3.1. The rod arrangement, grid f location, and the comparison of measured to predicted heat flux are shown on figures 3.10 through 3.12. The mean MR was 0.915 with a standard deviation of 0.843. i 3.1.4 Test Section ENC-6 This section represented a configuration typical of a 17x17 array with 0.360 diameter fuel. Twenty-four (24) heated rods and one ( unheated ceramic simulated guide tube were tested. The outside diameter of ( the heated rods was constant while the inside diameter was tapered to achieve a non-uniform axial heat flux. Characteristics of the test section and experimental < conditions are shown on Table 3.1. Rod arrangement, grid location, thermo-couple location, axial profile, and comparison between predicted and measured
- f. heat flux are shown on Figures 3.13 through 3.16. The mean POM3R was 0.995 with a standard deviation of 0.0749.
( l { ---
].
]
3-4 XN-W-621 (NP)(A) Revision 1
)
3.2 C. FIGHETTI AND D. REDDY DN8 DATA The experimental tests were conducted in the high pressure heat transfer facility at the Columbia Engineering Research Lchoratories of Colum-bia University. The data reported by Fighetti and Reddy( } includes results from major nuclear fuel vendors throughout the world. Several sections selected for analysis below included test sections u ing prototypic spacers and geome-try of Combustion Engineering fuel design and test sections with spacers and geometry prototypic of Westinghouse fuel design. 3.2.1 Combustion Engineering DNB Test Data Two test sections, each with 21 heated rods of 0.382 inch
' diameter are included in this analysis. Characteristics of the test sections and the experimental range of operating conditions are shown on Table 3.1.
One test section used a uniform axial while the other was a non-uniform sinusoidal axial power profile. The rod arrangement for test section CE-47 is
]
shown in Figure 3.17 while that for CE-59 is shown in Figure 3.18. The location of grid spacers (all non-mixing vane) and thermocouples are shown on Figure 3.19 while the non-uniform axial profile for CE-59 is show1 on Figure 3.20. Comparison of measured to predicted heat flux values are shown on Figures 3.21 and 3.22 for CE-47 and CE-59, respectively. The mean value of the ratio of predicted to measured heat flux for CE-47 was 1.028 with a standard deviation of 0.0741 while the mean for CE-59 was 1.023 with a stand- ) ard deviation of 0.0820. l s _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ - ~ _ _ . - - - -
s I s 3-5 XN-NF-621 (NP)(A) Revision 1 3.2.2 Westinghouse DN8 Test. Data Two test sections with 24 and 25 heated rods of 0.374 inch { diameter are shown in Figures 3.23 and 3.24. The spacer location and thermo-couple locations are shown in Figure 3.25 while the non-uniform sinusoidal ( axial profile is shown in Figure 3.26. Test section characteristics and experimental range of conditions are shown in Table 3.1. XNB predicted the critical heat flux over the range of conditions showi in Table 3.1. The mean DN8R for WH-162 was 0.992 while its su.'ndard deviation is 0.0845. For test { section WH-164, the mean is 0.950 with a standard deviation of' O.067/. The f predicted heat flux to measured heat flux is illustrated in Figures 3.27 and 3.28 for WH-162 and WH-164, respectively. - 3.3 ROSAL, ET AL TEST SECTIONS (I4) Four test sections of eight foot length are represented in this analysis. Rosal-4 represents a section in which the grids have no mixing ( vanes. Rosal-8 differs from Rosal-2 principally because of spacer pitch. Test section characteristics and operation conditions are shown on Table 3.1. Rod layout, grid location, and axial profiles are shown on Figures 3.29 through 3.31. Comparisons of measured to predicted heat flux is shown on Figures 3.32 through 3.35. Mean DNBR's were 0.976, 0.933, 0.970, 1.001 with standard deviation of 0.1118, 0.1043, 0.0987 for sections Rosal-2, Rosal-4, Rosal-7, and Rosal-8, respect ively. (. { l f .
1 3-6 XN-NF-621 (NP)(A) Revision 1 ) 1 3.4 DISCUSSION AND CONCLUSIONS OF ANALYSIS OF DNB TEST DATA The method to predict DNB heat flux described in Section 2.0 was used in the analysis of the data discussed in Section 3.0. The DNB prediction used a subchannel code to evaluate local flow conditions which are required as input to the Equations (2.1) through (2.10). These equationa correspond to
]
the XN8 correlation plus corrections for effects of non-uniform axial power distribution and georaetric parameters. Table 3.2 summarizes key statistical ) results for each section and overall. 3.4.1 Subchannel Mixing , Grid spacers promote subchannel mixing which reduces subchannel to subchannel enthalph gradients and tends to sweep vapor layers from the rod surface. This increases the DNB heat flux for a given set of ) fluid conditions. Depending on grid design, subchannel mixing can be a combination of forced diveston mixing and turbulent mixing. In the analysis of the data presented in this document, the calculation of mixing included flow diversion mixing (due to subchannel static pressure differences caused'by
)
grid spacer pressure losses) and turbulent mixing. forced diversion mixing was not . Included in ihe analysis. All subchannels of a given test section used the same grid spacer loss coefficient which corresponded to experimental-ly determined loss coefficients on grid spacers similar to those used in the test. The tutbulent mixing parameters used in the analysis of the ONB data were:
L (. - 3-7 XN-T -621 (NP)(A) Revision 1 l ( l where the basic tutbulent mixing equation 1s; W = SsG [ where: S = turbulent mixing parameter s = rod-to-rod spacing D = subchannel hydraulic diameter { G = subchannel mass velocity ( W = turbulent cross flow These values of Ss/D are based on experimental data (9) from a variety of fuel designs; Reference (9) verifles (.he above mixing relations to these data. 4 { l ( l
]
)
3-8 XN-NF-621(NP)(A) Revision 1
]
Table 3.1 Summary of Test Conditions Test Section ENC-1 ENC-2 ENC-3 ENC-4 ENC-5 Heated Lengtii (ft) 6 6 6 6 5.5 Axial Heat flux Distribution UNIFORM UN1f0RM UNIFORM UNIFORM UNIFORM
]
Hadial Power Distribution UNIFURM UNIFORM .95-1.10 .95-1.10 .95-1.08 Grid Design MG NV MV MV MV Hydraulle Diameter, Nominal Channel, anch .520 .520 .514 .514 .528 Rod 0.D., anch .413 .413 .421 .421 .424 ) Grid Spacing, inch 15.5 15.5 15.7 15.7 26.2 KLOSS**
]
Range of Experimental Parameters Pressure, psia 1500-2160 1500-2155 1500-2260 1500-2270 1745-2265 Inlet Temperature, *f 460-620 470-620 420-630 420-630 400-620 Inlet Avg. Mass Velocity, M1b/ht-ft2 1.0-2.56 1.00-2.53 1.0-2.77 1.0-2.72 .98-2.75
' Number of Data Points 28 24 73 80 59
- Mixing vane grida were on 22-inch pitch.
Simple support grids were halfway between MV grids. e* li
I L 3-9 XN-NF-621(NP)( A) Revision 1 (- Table 3.1 Summary -of Test Condit2cns (Continued) ( Test Section ENC-6 CE-47 CE-59 WH-162 WH-164 f Heated Length (ft) -12 12.5 12.5 14 14 Axial Heat Flux Distribution COSU UNIFORM COSU COSU COSU Hadial Power 01stributan .97-1.10 .97-1.14 96-1.20 .95-1.10- .94-1.10 ( Utad Design MV NV NV MV MV Hydtaulic Diametes, Nominal l Channel, Inch 0.5101 0.4714 0.4714 0.4635 0.4635 Hud 0.0., toch 0.360 0.382 0.382 0.374 0.374 f Grid Spacing, inch 20.56 14.30 14.30 22.0/11.0* 22.0/11.0* KLUSS
~
Range of Experimental Parameters ( Pressure, psia 1600-2400 1395-2405 1495-2415 1500-2425 1500-24'25 Inlet Temperature, *F 445-615 362-631 333-626 429-610 384-606 Inlet Avg. Mass Velocity, M1b/hr-ft2 0.9-3.1 0.9-4.0 0.9-4.0 0.9-3.1 0.9-3.1 Numbei of Data Points 62 96 89 53 53
- Hixing vane glids were on 22-Anch pitch.
f Simple support gtids were halfway between MV grids. l l { L - - - - - - - - - - - - - - - - -
I ] lI l 3-10 JCN-NF,-621 Revis10rr i(NP)( A) Table 3.1 Summary of Test Conditions (Continued) Tcst Section Hosal-2 Rosal-7
]
Rosal-4 Rosal-8 Heat.ed Length (ft) 8 14 8 8 Axial Heat flux Distribution U SINE U COSU U SINE U COSU . Radial Power Distribution No. Inner Rods - % Power 4-10C% 4-100% 4-100% 4-100% No. Outer Rods - % Power 12-83.1% 12-94% 12-83% 12-94%
)
Grid Design grids s./MY grids w/MV gritds w/o MV grids w/MV Hydesulle 01ameter, Nominal -
]
Channel, anch 0.507 0.507 0.507 0.507 ) Rod 0.0., anch 0.422 0.422 0.422 0.422 Grid Spacing, anch 20* 20* 20* 26* ) KLOSS Range of. Lxper'1 mental Parameters Pt essure, psia 1504-2410 1491-2105 1492-2148 1490-2432 Inlet Temperature, *f 466-627 479-580 481.5-603 478-626 Inlet Avg. Mass Veloc1Ly, ] Hib/h -tL2 2.02-3.58 2.07-3.63 1.50-3.62 2.02-3.61 - Numoer of Data Points 28 11 26 32
- Simple grad between indicated grid spacer.
J
L 3-11 XN-NF-621 (NP)(A) [ Revision 1 Table 3.2 Stat 2stical Summary [ Test Section Number Mean Standard Deviation { CL-47 96 1.028 0.0741 CE-59 89 1.023 0.0820 WH-64 53 0.950 0.0677 k WH-62 53 0.992 0.0845 ENC-6 62 0.995 0.0749 ( HUSAL-2 28 0.976 0.1118 HOSAL-4 26 0.933 0.0843 HOSAL-7 11 0.970 0.104) HOSAL-8 32 1.001 0.0987 LNC-1 28 1.0.29 0.1186 ENC-2 24 0.983 0.1084 ENC-3 73 0.939 0.0895 (- ENC-4 80- 0.985 0.1196 ENC-5 59 0.915 0.0843 ( TOTAL 714 0.984 0.0964 l (- [ [ [ _ - - _ -- -
]
l 3-12 XN-NF-621 ( NP)( A) Revision 1
]
m J 1 Ol.0 1.0 O.0 OO 1.0 1.0 ) Wl ( 1 ll W ll W i
;W }
1.0 jj lq Q WljCl'Q j
)
{ w }Iw w w {w ll Ii i H
]
l Q l l W ll Q lj O W
\\ O W WllWl
\
_ JL l lOl ] LOWER R00 NUtiBER PROVIDES R00 NUMBER SPECIFIED 1 IN DATA AS THE R0D INDICATING CHF , J
]
Figure 3.1 Test Assembly Geometry and Local Power Distribution for ENC 1 & 2 _ Test Assemblies im muni
k r L ( 3-13 XN-NF-621(NP)(A) Rev. 1 [ [ [ [ [ [ [ ( Figure 3.2 Spacer, Pressure Tap, and Thermocouple locations for ENC 1 Test [ f - _ - - - - - - - - - - - - - - - - - - - - -
r 3-14 XN-NF-621(NP)(A) - Rev. 1 m
)
w L Figure 3.3 Spacer, Pressure Tap and Thermocouple Locations for ENC 2 Test
]
1 4 m E ..
e N r 3-15 XN-NF-621 (NP)(A) L Revision 1 [. y - i i i i i i i i i i O Z ( - A LLJ
~ l o
[. - A X D
-.J LL-f
[ - eX G
- O d3 ts.
LLI I [ O DE k
- e -
og o [ e .o z e
*$s O
~
3,
=
%e H
[ r e ne m e occ c I ( - e - o o g 9 o at o ld m f D [- - e H i i i i i 9 1 i i e 0*f i e 0*T G'O 8*0 4*0 S*0 S*0 t*0 C*0 2*0 T*0 k 2*T T *T xnB .Ld3H 03HOSd3W [ [ [ L - - - -- _ --- _ ----- - - - -
J
,r 3-16 XN-NF-/>21 (NP)( A)
Revision 1 4 N s i i i i i i i i i i i - g a Z A W I o A X D LL 2 t-- e x G d3 4 to I g gi) e SE LL
, 9 oy a ee ,o z a - ee -
dW O J e o M H H se gg g g s@ e om n. a J
- o_
% i o o g 9
o in - o ld a E d D e H i i e i i i i i e i i 1 2 *T T *T O'T 6*0 8*0 40 3*0 S*0 t0 E*0 20 T*0 0*f XOli id3H 03BOSO3W l
- - _ _ - - - . _ . - - - - - - - - - - - - - - - - . - - - - - - - - - - _ _ - - - _ _ m
A 3-17 XN-NF-621 (NP)( A) ( Revision 1 I
.9505
.9518 l ( l.002 l { l.088 ; 0.00 j 1.002 l 1 2 3 4 5 l
I G 1.001 I i O 0.00 1 1 GX 1.094 ) I e1.002 I I sah 0.9518 l 6 17 18 19 6 [ I s i.093 I I e1.092 I Iel al e 1.003 l 1.001 l 0.953 l ( , 15 24 25 20 7 ( I 0.00 l.002 1.000 1 h ( 0.9988 l i 0.00 i 1 0.9518 l 23 22 { 8 [ j a1.003 I l a0.9547 l { e0.9525 l ( al ml 0.9517 ( 0.9518 13 12 11 10 9 [ TOP: TEST 1
. NUMBER OUTSIDE SHOWS R0D NUMBER Figure 3.6 Test Assembly Geometry and Local Power Distribution for ENC 3 & 4 Test
( Asserablies [ [ .
( 3-18 XN-NF-621(NP)(A) Rev. 1 I I I I I I l 1 I l l
- I I
k. Figure 3.7 Spacer, Pressure Tap and Thermocouple Locations for Tests ENC 3 & 4
-- r-- , r-- ,__ v <--,
v m e-, m_ r--, m r-m m r~-n i r----, r- -x 1 d i I I I I I I I I _J e4 / j _
- +16%
o A -
? e o -
e etDe x e 60 S-u.d - F-Es
-16%
1s - , @ Y l gm ms - e e - E e I w s - m s - o - {@
- =
s 5 Jn d -
- "5 z
- o. I i i i i i i i I i %
%.0 0.1 0.2 03 0.4 05 0.7 0.8 09 10 11 1.2 3 PREDICTED HEAT FLUX FIGURE 3.8 COMPARISON OF HEAT FLUX - ENC-3
i 3-20 XN-NF-621. (NP)( A) Revision 1 l n - s i i ; i i i i i i i - g l z A W g I o
- 0 -
A X 0 D b8 9 ee e se 43 x eI w I
%e =
u c 8 8e - S$9p e ' nti u-l e e e og o e e e 90 o I e 0 oG m l 5 H H 7 e O 90 ace cr ee ," 1 I O s g s n I e o b.
- % I
- - g D L9 H l i i i i i i i i i i i 9 k i 21 T *T O'T 6*0 8*0 40 30 S*0 t*0 E*0 20 T*0 0*f Xnid 103H 03MOS03W
[
.p.
^
[
.3 21 XN-W-621 (NP)(A) r Revision 1 L
l- [ (z ( ( 1 1 (: [; NUf18ER OUTSIDE SHOWS ROD NUMBER .
. Figure 3.10. Test Assembly Geometry and Local Power Distribution for ENC 5 Test Assembly
[ [ I _
u
'3-22 _XN-NF-621(NP)(A)
Rev. 1-
]
4 J c 1
~
)
1 1 i . Figure.3.ll; Spacer, Pressure Tap and Thermocouple
-Locations for ENC 5 Test Assembly
)-
~
ua u
x - ,
. m .;
4
- i i i i i i i i i i M
9
. S g _
e -
- x. e od -
ee - d r +16% gd
$b -
l u , 84o _ _ L m ! e a re e yd - e -
-16%
e d e J - g . - x x 4 af o - - p a s=_a. g - - . o* 1 1 I I I I l' i l i 1 3
*b.o 0.1 0.a o.3 c.4 0.5 o.s 0.7 0.8 09 1.0 11 12 -
PREDICTED HEAT FLUX FIGURE - _ - 3.12 COMPARISON OF HEAT FLUX - ENC.5 -
3-24 XN-NF-621(NP)(A) P,evision 1
)
1 1A.980 l l A 0.98 l l A0.98 l lA 0.98 } { A 0.98 l 1 J l 2 3 4 5
]
I OAAAA 0.98 l l 1.09 I i 1.09 l l 0.98 I I 0.98 I
]
16 17 18 19 6 f i 0.98 I I 0.98 li q
- g. i i 1.09 l l 0.98 j 15 24 9 g .D 20 7 25 1 0.98 l l 0.98
- I l 0.98 l {
l.09 l l 0.98 l 14 23 22 21 8 ) I 0.98 l j 0.98 l j 0.98 l j 0.98 l l 0.98 ' 13 12 11 10 9 u , NUMBER OUTSIDE SHOWS R00 NUMBER Figure 3.13 Test Assembly Geometry and Local Power Distribution for ENC 6 Test Assembly ]- ee __----a - - -
l l 1 l ll l s r y1* 3 xfM,@
. 23.b r
x ;.8: r r 0 1 r \
\
\
, % a
\ t
\ s e
m \ 8 T a 6 0 C N m . E r N a o f n
- o r .
a 6 i t 0 u b
- N i O r - . I a
L t r T / s A X i M D I X r 0 R a 4 e P w O o P Pn A o li R at
. f A a '
i c m E N I AS xe L 4 a 2 1
/ 0 3 L f e A / r
. E / R u D / g I / i
/ F
/
6 o 4 2 0 8 6 4 2 Q, 1 1 1 1 0 0 0 0 Q
@ g <w" $.x< .
! ll l
3-26 XN-NF-621(NP)(A) Rev. 1 '
)
)
J s u t Figure 3.15 Spacer, Pressure Tap and Thermocouple Location for ENC 6 Test Assembly 1 J 1
H; I r L 3-27 XN-NF-621 (NP)(A) (! Revision 1 ( [ ~ . i i i i i i i i i i - g w Z (i - e - A LU i (
- o i -
4 X
- .- D
.J
(, - e - d k n
.g a dd (i E
[:
- %e E, -
e og u. a e e .o z
- 0 -
dS' O u m [ M O H H g
- 7 -
*8 c a:
m c r (i i O E
," 1 I
e - d 8 (l - - 9 b a e; ( - - d . to a E ( - - d D e
- H i e i i i e i e i e i k 21 tT ot so 8o co so so to co ao To of
. XG1d id3H 03HOS03H i
e t il
3-28 XN-NF-621 (NP)(A) Revision 1
]
.123 l .977 14 15 16 17 6 13 20 21 18 7
.976 .976 .974 -
12 jg 8 > 0.0 1 0 9 ar - O.97 NUMBER OUTSIDE SHOWS R0D NUMBER Figure 3.17 Test Assembly Geometry and Local Power Distribution For CE-47 Test J Section
e L 3-29 XN NF-621 (NP)(A) p Revision 1 L [ .123
.382
- '506 ; I l -4 4-k 9r
.965 I .968 .975 .968 .971 l
. 1 2 3 4 5 14 15 16 17 6 b 13 20 1 18 7 t m
[ , f u 1 0 9 0.98 ( NUMBER OUTSIDE SHOWS ROD NUMBER- , Figure 3.18 Test Assembly Geometry and Local Power Distribution For CE-59 Test Section t
3-30 XN-PE A?1 (NP)(A) GRID THERM 0COUPl.E Rev. 1 LOCATION LOCATION 144.61 143.16 130.41 - 129.16 116.21 - 114.95 102.01 - 100.76 87.81 - 86.56 73.61 - 72.36 59.41' - 45.21 - 31.01 - 16.81 2.61 INLET Figure 3.19 Spacer and Thermocouple Locations For CE-59 Test Section and Spacer locations For CE-47 Test Section
i llll uO- xzk,md o 2wb o r
- 2 E oa w n_
0
- m. - - - - - -
1 m_ 8 0 r 9 5 E C r r 6 o
. ' f 0 n i
o _ L t r_ .
' / u X b i
r t
- 4 s r_ i 0 D n
ro ei wt c . ' oc Pe S l at 2 is
. ' xe r 0 AT 0
2
. 8 .
m 3 e 0 r
- - - - - - u g
n 0 6 4 2 0 8 6 4 2 0 i F 1 1 1 1 0 0 0 . 0 0 . az u$a. J< 0<
- l l ll
3-32 XN-NF-621 (NP)(A) Revision 1 1' I
, , , , , , , , , , , 2 : l m
1 0
\
_ \ ~ 9
~
x
\,x
,\ ~
3l m l--
\, G x
s 23u. s w& LL
~
g _ g$ o
- 1 9e or 5l y)
N M H i oN mo e C i e e - ddo. o_ l
$+ 1
- o s o n
d -
~
- d .
La T 3
- c. g
, , , , , , , , , , , e '
2 *I T *I 0*T 6*0 8*0 4*0 30 S*0 t*0 E*0 20 T*0 0*0 xnid id3H 0380Sd3W {
m r- _ nw t w N d I I I I I I I I I /l I /
/ /
l ,' -
/
o -
/
a -
/
'?
o _ - O
- x. eQ .
l 3 ~ he l ~ l
' e g ES wo e - !
2 +16% g : 04 - eg e
- 0 i
u 09 I O -16%
~
i. s - - i n i d - - i
* =E 1 N 2' -
o - - f o -
/ -
E9 , 9 i i i i e i e i i i i %
%.0 01 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 11 12 3 PREDICTED HEAT FLUX FIGURE ,
3.22 COMPARISON OF HEAT FLUX - CE59
3-34 XN- -621 (NP)(A)
.374
)
.100 e y H 4 6 7
2 4 6 7 8 9 6 5 4 25 20 I 14 3 22 1 ) y 3 12 1 0 9 NUMBER OUTSIDE SHOWS R0D NUMBER Figure 3.23 Test Assembly Geometry and Local Power Distribution For W-162 Test Section
L l-35 ( XN-NF.671 (NP)( A) Revision 1 ( (
.374 .100 0 496 3 g g g
, r 1 2 3 4 5 16 17 18 19 6 4 15 24 25 20 7
( [ 14 23 22 21 y 88888 13 12 11 10 9
" S""
**!*1^!i""Jlsoti:T0"f649est Section
i ! 3-36 XN-NF-621 (NP)( A) Revision 1 GRID THERM 0 COUPLE LOCATION LOCATION s 157 MV - 146 SS - 135 MV - TC 114 124 SS - - TC 123.5 113 MV - - TC 112.5 102 SS -
-- TC 101.5 f
91 MV -
- TC 90 80 SS -
69 MV - SP SS -
)
47 MV - 36 SS - 25 MV - 14 SS - 3 MV - INLET
- Figure 3.25 Spacer and Thermocouple Locations for Test Sections W-162 and W-164; MV = Mixing Vane Grid; SS = Simple [
Support and TC = Thermocouple. [ Distances From Start of Heated Length.
@ 9 1-37 XN-NF-621 ( NP )( A)
Revision 1 { C ( u a y u u -T ' a (- .
~ -
CO d i 3 s
. - 2
(
- E o 3 e
( bS
.. - e x
a o. g M W *J e e o e a. - E@ ( . . a, T [-
. - 9 o
2 g E . I a . n . 'a
- - - - C O C C DNI'AV3d 7VIXV 0
4 _ - . _ _ _ _ _ - . . _ - -
3-38 XN-NF-621 (NP)( A) Revision 1 w* nj ] d CD i i i i i I i i I a l 1 a I J
- 4 3 I
o 4 X D _J a e LL 2' 5 x H o G e - W W I e >- I.J e d O ) a
% =a e+ e w Z g
.. "? (D oa s e W x G
i Q.
- e -
o O 1 e U e _ g 4 % a ; s &
- D
(.D i i i i e i i i i i 21 T1 oT so 8o 40 so to to 20 To of l XOld id3H 03HOSd3W h {
s .m_ - _v v n r s- w , N
" I i I I I i I i i I /
e4 R- -
/
R o _
/ _
s9o ' / ' g gg /
+16%
gd - O - 2
%e 8 om O y yd - -
g
@ ON -16%
GW yd - - e i @ o _ _ C O o _ _ N d - - Y p 4 o _ _ - 5 o* 1 1 1 l l l l l 1 I f Q 0.0 0.1 0.2 0.3 0.4 0.5 06 0.7 0.8 0.9 10 1.1 1.E - PREDICTED HERT FLUX FIGURE - 3.28 COMPARISON OF HEAT FLUX - WH164
3 40 XN-NF4621 (NP)(A) Revision 1
.153
.555 .422
.95
.98 lh .95
.98
.95
.98
.95 1.15 1.1 .95
.98 1.05 1.05 .98 12 13 14 5
.95 1.15 1.15 .95
.98 1.05 I 1.05 ) .98 )
11 16 15 6
.95 .95 .95 .95 I I
.98 .98 .98 .98 10 9 7
' ^
2.899 UPPER POWER FOR TEST SECTIONS 2,6 - LOWER POWER FOR TEST SECTIONS 7,8. NUMBER OUTSIDE SHOWS R00 NUMBEP,
]
Figure 3.29 Test Section Geometry and Locil Power Distributions for ROSAL 2, 6, 7 and 8 f Test Assemblies- {
s i ? s [ 3'41* XN-NF-621 (NP)(A) CUSIit u Revision 1 U SInt U ( 5 MV - - TC 1 6"
- TC 2 9"
- TC 3 12 III MV ~- TC 1 12"
(- - TC 2 15" 15 SS -- TC 4 16"
- TC 3 18"
- TC 5 20"
( 21 SS - - TC 4 22" 25 MV - - TC 6 26" 314 MV - - TC 5 32" 35k SS -
- TC 6 38" 41% SS -
45% MV - 51k MV - 55k SS - 61% SS - 65k MV - ( 71 MV - 75% SS - 81% SS - 85 MV - 91% MV - . INLET INLET ( Figure 3.30 Spacer and Thcrnoccuplc Locations for ROSAL 2, 6, 7 and 8 Test Sections; MV = Mixing Vane Grid; SS = Simple Support and TC = Thermcouple Location. Distances From End of Heated Length.
! l l
\\
uI" . Eke.O.%23S~ 4 S
- S.
U E N I S 0 U 1 a 8 t
. , a 0 D L
A S O R r o 6 f 0 s n i o t
, L u
- / b X i r
t 4 s
, i 0 D r
e w
, o P
l U a 2 i H E , x T N 0 A G I N S E O L C 1 3 D E 3 T t A f e E 0 r H 8 . u 0 g E i L F D N . U B T \ S E T W 0 8 6 4 2 0 8 6 4 2 2 1 1 1 1 1 0 0 0 0
! B 2" g I<
1ll llll lll ,'
- I 3-43 XN-NF-621 (NP)( A) Revision 1 (
~
4 d ( 1 I I l 4 I I I I I - CI e V. _ _ a C i
- _ e, N
O e E e , e - d E x F e* - d3 C e ee a
=
w k 2 e i _c. e e
=y g o, o
- _ g$ 2 g O C e H (/)
7 $
=x h
[ - A ct: C w G-I
, C i U o
- i. - _
s N N o A
-4 b d g;
[ G i ! i i i e i e i i i 9 H 2*T T *T O'T 6*0 8*0 40 S*0 S*0 ?*0 C*0 20 T*0 0*f k XO"13 103H 03BOSd3W
3-44 XN- % 621 (NP)(A) Revision 1 N e i i i I i I I I I d NI s - - G (n o g o i A W
\ , D
. __I o g "5
- e e -
os G
- h. L.LJ O r I 5 ~
LL. T o ao gW z u O a o M e 9w H 7 \ s s - om & d e n s N R s A a LJ f & D (.D i e i i i i i i i i i H ao co so so co ao of k 21 TT oT so to To xnb 1.83H 03BnSO3W {
m_ r _
,w 4
d
- : : i i ,
i T
,/ -
o A
/
/
e d - g l x co O O l 3d - O - l $ +16% O e
.- ~
ag _ e _ Y e w Q2 b yd - e - - o -16% m Gin yd - - i, _ _ n / s _ _ my 4 o - - 0 y%, a / sn se s _ _ -
- o. I l l I I I I i i i l 3 3.0 0.1 0.E 0.3 0.4 0.5 0.s 0.7 0.8 09 10 11 12 -
PREDICTED HEAT FLUX FIGURE .3 34 COMPARISON OF HEAT FLUX - ROSAL.7 ,
i d' 1 I I I I I I I I I I a A - 9- - m
~
~
ee e 5"o - 9 e -
- s u.
+16%
ee Hh as - ae - 1 E ! om @e y l yd e e -
. -16%
can yd - i o _ n s - n d - 5. ? EM 1 aa o - J g 3 9 i e i i e i i i a i E
*b.0 01 02 0.3 0.4 0.5 0.s 0.7 08 09 10 11 12 PREDICTED HEAT FLUX FIGURE . 3.35 COMPARISON OF HEAT FIIlX - ROROI R - - - _ - - - m
i r 4-1 XN-NF-621 (NP)(A) Revision 1 4.0 STATISTICAL EVALUATION The experimental data were used to determine the departure from nucleate boiling tatio ctiterion, u, which satisfies the following statistical state-( ment: With 95*. confidence, at least 95!i of the population of DNBR are less than u. This is referred to as a 95/95 toletance statement. Calculational steps for computing u are outlined below. from the esti-mates of means and standard deviations for the individual test sections an estimate of' the overall mean (a weighted average)* and the overall standard deviation, o y , are determined. Two variance components are calculated: (1) The within test section variance, o ,2 [ (2) The between test section variance, o B lhe Lolt ratice crit erval is constructed by methods given by Weissberg and Bet.t y . ( 10) The interval requires knowledge of the cegtees of freedom assoc 1-aled with 0 (= o g + og h, and the effective sample size, N, for the esti-7 mate of the weighted average. The degrees of freedom for ei are f und by Satterthwaite's fo'rmula(
, to be 396. The e f fective sample size, N, is the number of observations t equit ed to be selected at random from the population to give an estimate of
", hav ing a variance of .00010586, wnich is the variance of the weighted avetage. The value of N la found by solving: Ihe wesghted average au based on the numbet of data points of each test section und the telutive sizes of the variance between and within test Secllofis.
4-2 XN-NF-621 (NP)(A) Revision 1 2
'T = .00010586 N
The limit for u is then derived from u = p + Koi where K 18 given in Reference 13. Therefore, the tolerance statement becomes: With 95% confidence at least 95% of the DNBR (predicted to measured DNB heat flux) values are less than 1.16 for all the data analyzed. ,
5-1 XN-NF-621 (NP)(A) Revisicn 1
5.0 REFERENCES
L (1) K.P. Galbraith and T.W. Patten, "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady State and Transient c Core Operation," XN-75-21. b (2) R.B. Macduff and T.W. Patten, "DN8 Test Data R epo r t ,'- XN-NF-81-80, Revision 1, January 15, 1982. (3) C. Fighetti and D. Reddy, " Compilation of Critteal Heat Flux Data Taken at Columbia University," to be published as EPRI Report EPRI t)roject RP-813. (4) T.W. Patten, " Test Specification for a 17x17 DNB Test Program," XN-T-P62,659, July 1980. (5) H.G. Hartman, " Analysis of the Palisades DNB and Mixing Tests," XN 65(P), January 1976. (6) C.E. Leach, " Departure From Nucleate Boiling Test Specification," XN-L S30207, Revision 1.
~
( (7) J. Yates, "H.B. Robinson Departure From Nucleate Boiling Test Anslysis and Results," XN-75-45, October 20, 1975. (8) K.P. Galbraith, T.W. Patten, J.L. Jaech and M. McMullen, " Definition and L Justification of Exxon Nuclear Company DN8 Correlation for PWR's," XN-75-48, Octobei 1975. ( (9) R.B. Macduff, " Turbulent Mixing in Rod Bundles," XN-NF-81-73, October 1981. (10) A. Weissberg and G.H. Beatty, " Tables of Tolerance Limit Factors for Normal Distributions," Technometrics 2 No. 4 483-500, November 1960. (11) F.E. Satterthwaite, "An Approximate Distribution of Estimate of Variance Components," Biometrics Bulletin 2, 1946. (12) L.S. Tong, " Boiling Crisis and Critical Heat Flux," AEC Critical Review Series, 1972. (13) D. B. Owen, " Factors for One-Sided Tolerance Limits," S"R-607, March 1963. (14) E.R. Rosal, J. Cermak, L. Tong, J. Casterline, S. Kokolis, B. Matzner, "High Pressure Rod Bundle DNB Data With Axially Non-Uniform Heat Flux," ( Nuclear Engineering and Design, 31 (1974) pp.1-20. t
A-1 XN-NF-621 (NP)(A) Revision 1 DATA StNMAflY FOR TEST SECTION _ ENC _1 ( PRESSURE MASS FLUX INLET SUBC00 LING LOCAL ENTHALPY HEAT FLUX MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDNBR 1.065 1.018 1.006 1.202 1.131 1.128 ( 1.149 1.082 1.046 { 1.169 1.212 1.235 1.227 1.110 1.078 0.957 0.949 0.968 [ 0.966 L 0.955 0.906 0.896 0.896 0.885 0.942 ( 0.890 L 0.887 0.872 . ( l L__
A-2 XN-NF-621 (N?)(A) Revision 1 DATA
SUMMARY
FOR TEST SECTION - ENC-2 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 kBTU/MRFT2 MDNBR 1.035 0.995 1.068 1.137 1.064 1.027 1.124 1.082 1.060 1.185 1.095 1.068 1.031 0.883 0.685 0.882 0.869 0.868 0.866 0.883 0.869 0.838 0.931 0.867
h I' ~ r A-3 XN-NF-621 (NP)(A) L Revision 1 DATA
SUMMARY
FOR TEST SECTION - ENC _3 INLET LOCAL lEAT FLUX { Par.SSURE ' MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA- -MLB/HRFT2 BTU /LBM- BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MONBR 0.991 1.199 1.158 f 1.062 1.024 1.013 {- 0.952 0.975 0.994 1.116 0.979
, 0.939 0.912 1.067 1.027 1.000 0.946 0.971 0.987 0.945 0.963 1.020 1.029
[ 0.975 1.016 0.923 0.889 f- 0.877 0.808 - 0.992 0.919 0.808 0.803 L 0.893 f 0.856 0.816 0.863 ( 0.834 0.895
- 0.867 l -- _ _ --
A-4 XN-NF -621 (NP)(A) Revision 1 DATA
SUMMARY
FOR TEST SECTION - ENC-3 ~ INLET LOCAL HEAT FLUX PRESSUHE' MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASL PSIA MLB/HRFT2 BTU /LBM BTU /LBM teTU/HRF T2 MBTU/HRFT2 MONSR 0.905 0.881 0.922 0.888 0.869 0.858 0.905 U.895 0.954 1.077 1.011 0.985 1.210 0.841 0.929 0.842 0.897 0.864 0.867 0.910 0.882
-0.867 0.857 1.014 0.933 0.901 0.886 0.872 -
0.743 1.008 0.950
\
0.911 0.932 {
s ?^ L-A-5 XN.NF-621 (NP)(A) ( Revision 1 DATA SINMARY FOR TEST SECTION - ENC-4 {. INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SU8C00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MTU/HRFT2 M8TU/H;1FT2 M02R 0.915 1.176 1.152 { 0.999 0.948 1.400 (; 1.268 1.083 1.410 [- 1.162 ( 1.051 1.012 0.997
- f. 0.986 0.941 0.882 1.193
{ 1.048 0.787 y 1.155 l 1.099 1.031 0.938 1.161 ( 1.087 0.917 0.876 1.244 1.095 . 0.961 1.013 f 0.938 0.900 0.917 0.942 0.867 0.900 l 0.872 0.905 0.938 3
A-6 XN-NF-621 (NP)(A) ] Revision 1 J DATA SlHMARY FOR TEST SECTION - ENC-4 PRESSURE MASS FLUX INLET SUBC00 LING LOCAL ENTHALPY HEAT FLUX MEASURED PREDICTED [ CASE PSIA ML8/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MOPER 0.919 1.092 0.990 0.968 0.906 0.892 0.051 0.921 , 0.886 0.852 0.950 0.973 1.029 0.885 1.030 0.944 0.939 0.923 0.890 0.929 0.948 0.992 1.042 0.909 0.932 0.931 0.935 1.013 0.918 - 0.918 1.065 0.784 0.929 1.030 0.921 0.936 0.907 0.872 ) 0.962 1.021 l
s. I L ( A-7 XN-NF-621(NP)(A) . Revision 1 DATA SLNMARY FOR TEST SECTION - ENC-5 l INLET LOCAL EAT FLUX PRESSURE MASS FLUX SU8C00 LING ENTHALPY MEASURED PREDICTED CASE PSIA H.B/HRFT2 8TU/L8M BTU /L8M M8TU/HRFT2 MBTU/HRFT2 MD.'SR 0.884 0.908 f 0.816 0.857 1.176 ( 0.864 0.866 0.846 [ 1.201 l ! 0.954 0.830 0.800 f~ 0.855 1.054 0.895 (, 0.807 0.926 0.878 0.865 ( 0.801 0.840 0.891 0.965 { 0.830 0.917 1.012 ( 0.880 0.884 - 0.891 1.052 1.049 0.924 [. 0.846 1 0.882 0.858 0.856 f 0.832 0.811 1.015 0.963 I l
A-8 XN-Pf-621 (NP)(A) Revision 1 DAI A SLNMARY FOR TEST SECTION - ENC-5 INLET L0t.f . HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDMSR 0.959 0.982 1.001 1.023 0.884 0.974 0.925 0.989 0.985 0.890 0.910 0.873 0.899 0.976 O.956 0.885 0.864 0.860 0.874 l
- d. '
A-9 XN-NF-621 (NP)(A) L Revision 1 DATA SINMARY FOR TEST SECTION ENC-6 f' - INLET LOCAL HEAT FLUX { PRESSURE M SS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED 2 CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 HDM3R 1.035 0.895 0.990 1.070 0.988 0.896 1.105 1.087 0.957 1.059 1.048 0.953 0.943 (- 0.995
. 0.959 0.963
{' 1.082 1.001 1.004 1.037 f 1.122 0.988 0.988 0.865 1.191 0.860 0.907
> 0.927 0.991
[ 0.921 l 0.866 0.912 1 0.960 f- 0.969 0.946 0.967 0.972 { 1.007 0.970 0.985 i
a... .- A-10 XN.NF-621 (NP)( A) Revision 1 DATA
SUMMARY
FOR TEST SECTION - ENC-6 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SU8C00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 @NBR 0.989 0.915 1.149 0.970 0.949 0.921 0.936 1.031 1.047 0.941 0.989 0.992 0.968 r 0.983 1.064 { 1.119 0.959 1.001 1.049 1.211 1.046 1.092
L I' B-1 XN-NF-621 (NP)( A) L Revision 1 i ( DATA SINMARY FOR TEST SECil0N . ROSAl_2 INLET LOCAL HEAT FLUX
~ PRESSURE MASS FLUX SU8C00 LING- ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDMBR 0.726 r 0.866 l 0.915 1.134 1.072 1.198 0.857 0.815 0.995 0.984 O.987 0.972
- f. 0.976 0.965 0.931 0.833 0.917 0.891 0.910 0.982 0.911 L 1.002 1.099 0.996 1
1.003 ( 1.131 (. 1.086 1.177 . i
)
)
B-2 XN NF~621 (NP)(A) Revision 1 DATA
SUMMARY
FOR TEST SECTION - ROSAL_4 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE - PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDNiiri __________________________________________________________________________________ = 1.074 0.892 0.873 0.976 0.986 0.907 1.037 0.899 1.074 0.968 0.886 0.865 0.til7 0.918 0.932 0.921 0.915 0.878 0.906 0.821 0.897 0.884 1.208 0.886 0.896 0.915 l
q: f L L: B-3 XNM-62.1 (NP)( A) Revision 1 I- ' DATA SlNMARY FOR TEST SECTION . ROSAL-7 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA. MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDNBR ('.- _________ ____________ ...________ .. ..__________________________________________ 0.934 O.883 {' 0.845-1.121 1.107 [ t 0.863 0.982 1.135 0.931 f 0.960 0.919 f .. { { i (
B-4 XN-NF-621 (NP)(A) Revision 1 DATA
SUMMARY
FOR TEST SECTION - ROSAL-8 INLET LOCAL HEAT FLUX PRESSURE t%SS FLUX SUBC00 LING ENTHALPY' MEASURED PREDICTED . CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MON 3R 0.929 0.896 0.971 1.029 1.139 0.917 1.049 0.935 0.912 0.899 0.917 0.855 0.878 0.904 0.894 0.921 0.967 0.893 0.919 1.187 1.070 1.140 1.067 1.177 1.039 1.089 1.030 ( 1.044 - ( 1.073 1.100-1.107 1.105 l l
4 5 C-1 XN-NF-621 (NP)(A) L Revision 1 DATA SlMMARY FOR TEST SECTION - CE-47 ( [- INLET LOCAL HEAT FLUX l PRESSURE . MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE- PSIA MLB/HRFT2 BTU /LBM BTU /LBM M8TU/HRFT2 MBTU/HRFT2 M0hBR
; 1.120 1.024
( 1.131 1 1.047 1.111 1.032 1.009 1.122 1.134 1.090 1.069 1.009 1.019 1.006 0.913 0.953 0.952 0.927 1.120 0.879 { 1.009 1.085 1.083 1.043 0.983 0.971 1.217 1.137 ' i 1.161 1.181 1.023 1.121 1.060 1.033 1.006 1.000 1.050 1.003 1.010 0.984
l-
)
C-2 XN-NF-621 (NP)(A) Revision 1 DATA
SUMMARY
FOR TEST SECTION - CE-47 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUSC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDM3R 1.079 1.091 1.085 1.069 1.048 1.057 1.046 1.009 1.141 1.059 1.052 1.034 0.982 0.971 1.098 1.016
-0.943 1.013 0.959 1.032 0.994 0.993 0.985 1.001 (
1.000 [ 0.971 1.007 0.967 , 0.889 1.126 1.131 ( 1.084 { 1.007 0.935 , 1.069 1.110 0.941 0.909 0.943 0.845
s. L ( ' C-3 XN.NF-621 (NP)(A) Revision i DATA SLJ4 MARY FOR TEST SECTION - CE-47 INLET' ( PRESSURE MASS FLUX SU8C00 LING LOCAL ENTHALPY HEAT FLUX MEASURED PREDICTED CASE PSIA' MLB/HRFT2 BTU /LBM BTU /LBM HSTU/HRFT2 MBTU/HRFT2 MDr6R 0.936 0.964 1.013 0.878 0.890 1.024 ( 1.085 1.166 r 1.127 l 1.050 1.027 1.103 1.006 (' 0.995 0.947 1.001
~
I
)
I: C-4 XN-NF.621 (NP)(A) Revision 1 DATA SLNMARY FOR TEST SECTION - CE-59 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM M8TU/HRFT2 MBTU/HRFT2 MDBBR 1.037 0.958 1.051 1.050 1.076 1.069 1.010 0.977 1.012 0.899 1.019 0.945 1.003 1.123 1.100 { 1.114 ( 1.114 1.079 1.054 1.050 1.079 1.040 1.076 0.992 1.045 1.043 1.075 0.860 1.040 , 1.063 1.113 1.142 1.062 0.859 0.912 0.966 1.077 1.171 1.063 1.057 .
{ l-(J C-5 XN_NF_621' (NP)( A) Revision 1 {: DATA SINMARY FOR TEST SECI10N - CE-59 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUSC00 LING ENTHALPY MEASURED PREDICTED { CASE. PSIA MLB/;RFT2 BTU /LBM BTU /LBM MTU/HRF72 MBTU/HRFT2 MDNBR L 0.842 1.037 0.957 f; 0.987 0.975 0.978 O.992 0.970 1.164 1.136 1.006 f' ~ 1.034 L 0.971 1.007 1.121 1.113 1.195 0.899 0.876 0.914 1.041 1.077 0.978 I 0.789 (- 1.037 0.973 g 0.892 0.932
; 1.200 ,
1.046 i 1.017 ! 1.049 1.020 1.127 1.021 0.903 1.109 1.098 1.055 0.994 4 t e* +g,-t- ,-- ,----er,- p-,,-p-%---9,,,-,+,w,.--,-,----a..,p. ,,,g., g pg9y9. ..%g,g .g . 9p,_.,7,
C-6 XN_NF_621 (NP)(A) Revision 1 DATA
SUMMARY
FOR TEST SECTION - CE_59 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDER 1.039 1.169 0.961 0.925 1.063 0.957 0.998 1.028 0.958 l 1 l 1 I l I i i
L I' [. C-7 XN_NF 521 (NP)(A) Revision 1 UATA
SUMMARY
FOR TEST SECTION . WH_162 g. [ INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM M3 TU/HRF T2 MBTU/HRFT2 HDNBR ( __________________________________________________________________________________ 0.818 1.016 ( 0.944 0.945 0.979 f 0.988 0.925 0.948 0.954 ( - 0.991 0.969 ( 1.024 1 0.973 0.974 1.049 0.968 0.926 0.989 0.950 { 0.965 1.015 0.996 1.082 1.014 1.100 0.876 0.987 0.903 - 0.848 1.063 0.919 1.022 1.027 0.942 0.985 1.110 0.950 0.096 0.896 1.067
. g.
. 2n.Y%
C-8 XN.NF-621 (NP)(A) _ Revisioa 1 DATA
SUMMARY
FOR TEST SECTION - WH-162 INLET LOCAL HEAT FLUX ] PRESSURE MASS FLUX SUSC00 LING .ENTHALPY MEASURED PREDICTED CASE' PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDNBR 1.182 1.028 1.087 1.120 0.985 1.103 0.789 0.945 1.122 1.091 0.808 1.104 1.115 f 1 i l
u f f C-9 XN-NF-621(NP)(A) Revision 1 ( DATA
SUMMARY
FOR TEST SECTION - WH-164 f INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MONBR { 0.785 0.898 0.914 0.919 0.877 f- 0.861 0.890 0.969 0.929 (- 0.922 O.909 0.941 0.993 0.937 0.955 0.943 1.010 0.846 0.956 0.877 0.884 0.832 0.880
} 0.858 }
0.980 0.906 0.976 0.996
- 1.007 0.994 0.932 0.940 0.981 0.858 0.991 0.951 1.001 1.057 0.975 1.025
'e
\
C-10 XN-NF-621 (NP)( A) Revision 1 DATA SINMARY FOR TEST SECTInN - WH-164 INLET LOCAL HEAT FLUX PRESSURE MASS FLUX SUBC00 LING ENTHALPY MEASURED PREDICTED CASE PSIA MLB/HRFT2 BTU /LBM BTU /LBM MBTU/HRFT2 MBTU/HRFT2 MDNBR 1.046 0.905 0.981 0.949 1.009 1.002 0.944 1.105 1.057 1.071 1.050 1.021 0.869 1
L I-I XN-NF-621(NP)(A) Revision 1 Issue Date 10/21/83 L f'. EXXON NUCLEAR DNB CORRELATION FOR PWR FUEL DESIGNS
-DISTRIBUTION J. C. Chandler R..B. Macduff J. N. Morgan f J. C. Chandler /NRC (16)
Document Control (5) ( {
.. - _ _ - - - -}}