ML050180257
| ML050180257 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Surry, North Anna  |
| Issue date: | 01/13/2005 |
| From: | Grecheck E Dominion Nuclear Connecticut, Dominion Resources, Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 05-020, DOM-NAF-2 | |
| Download: ML050180257 (33) | |
Text
k Dominion January 13, 2005 United States Nuclear Regulatory Commission Serial No.05-020 Attention: Document Control Desk NL&OS/ETS Washington, D.C. 20555 Docket Nos. 50-280/281 50-338/339 50- 336/423 License Nos. DPR-32/37 N PF-4/7 D PR-65/N P F-49 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
DOMINION NUCLEAR CONNECTICUT, INC (DNC)
NORTH ANNA AND SURRY POWER STATIONS UNITS 1 AND 2 MILLSTONE POWER STATION UNITS 2 AND 3 REQUEST FOR APPROVAL OF APPENDIX B OF TOPICAL REPORT DOM-NAF-2 QUALIFICATION OF THE WESTINGHOUSE WRB-1 CHF CORRELATION IN THE DOMINION VIPRE-D COMPUTER CODE VIPRE is a core thermal-hydraulics computer code developed by EPRl and approved by the NRC, which is currently in wide use throughout the nuclear industry. VIPRE-D is the Dominion version of VIPRE, which has been enhanced by the addition of several vendor specific CHF correlations.
Dominion has validated VIPRE-D through extensive code benchmark calculations. In addition, the accuracy of VIPRE-D has been demonstrated through comparisons with other NRC-approved methodologies.
In a September 30, 2004 letter (Serial No.04-606), Dominion submitted Topical Report DOM-NAF-2, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, and Appendix A to the Topical Report DOM-NAF-2, Qualification of the F-ANP BWU CHF Correlations in the Dominion VI PRE-D Computer Code, for NRC review and approval.
Continuing with the modular approach to Topical Report DOM-NAF-2, and as discussed during the public meeting held between the NRC and Dominion on August 4, 2004, Dominion is now submitting Appendix B to this Topical Report, Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code, for NRC review and approval. Appendix 6, which is provided in Attachment 1 to this letter, documents the qualification of the Westinghouse WRB-1 CHF Correlation with the VIPRE-D code and the code/correlation DNBR design limits.
Dominion continues to request the approval of the generic application of this topical report and the associated appendices. Plant specific applications of this topical report, including
applicable appendixes, will be submitted accordance with Section 2.1 of DOM-NAF-2.
Serial No.05-020 Dockets 50-280/281, 50-338/339, 50-336/423 VlPRE D Computer Code Page2 of 3 to the NRC for review and approval, in If you have further questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.
Very truly yours, Eugene S. Grecheck Vice President - Nuclear Support Services Virginia Electric and Power Company and Dominion Nuclear Connecticut, Inc Attachment Commitments made in this letter: None
Serial No.05-020 Dockets 50-280/281, 50-338/339, 50-336/423 VlPRE D Computer Code Page 3 of 3 cc:
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW - Suite 23T85 Atlanta, Georgia 30303 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406-1 41 5 Mr. M. S. King (w/o Att.)
NRC Senior Resident Inspector North Anna Power Station Mr. N. P. Garrett(w/o Att.)
NRC Senior Resident Inspector Surry Power Station Mr. S. M. Schneider (w/o Att.)
NRC Senior Resident Inspector Millstone Power Station Mr. V. Nerses NRC Senior Project Manager - Millstone Unit 2 U. S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852 Mr. G. F. Wunder NRC Project Manager - Millstone Unit 3 U. S. Nuclear Regulatory Commission One White Flint North 11 555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852 Mr. Stephen R. Monarque NRC Project Manager - Surry and North Anna U. S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 8-HI 2 Rockville, Maryland 20 Appendix B to Topical Report DOM-NAF-2 QUALIFICATION OF THE WESTINGHOUSE WRB-1 CHF CORRELATION IN THE DOMINION VIPRE-D COMPUTER CODE Virginia Electric and Power Company (Dominion)
Dominion Nuclear Connecticut, INC (DNC)
DOM-NAF-2, Rev. 0.0 APPENDIX B Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code NUCLEAR ANALYSIS AND FUEL DEPARTMENT DOMINION RICHMOND, VI RG I N I A December, 2004 Prepared by:
Sean M. Blair Rosa M. Bilbao y Leon Reviewed by:
Dana M. Knee Recommended for Approval:
Supervisor, Nuclear Safety Analysis Director, Nuclear Analysis and Fuel
CLASSIFICATION/DISCLAlMER The data, information, analytical techniques, and conclusions in this report have been prepared solely for use by Dominion (the Company), and they may not be appropriate for use in situations other than those for which they are specifically prepared. The Company therefore makes no claim or warranty whatsoever, expressed or implied, as to their accuracy, usefulness, or applicability. In particular, THE COMPANY MAKES NO WARRANTY OF MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USAGE OR TRADE, with respect to this report or any of the data, information, analytical techniques, or conclusions in it. By making this report available, the Company does not authorize its use by others, and any such use is expressly forbidden except with the prior written approval of the Company. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall the Company be liable, under any legal theory whatsoever (whether contract, tort, warranty, or strict or absolute liability), for any property damage, mental or physical injury or death, loss of use of property, or other damage resulting from or arising out of the use, authorized or unauthorized, of this report ABSTRACT This Appendix documents Dominions qualification of the Westinghouse WRB-1 CHF correlation with the VIPRE-D code. This qualification was performed against the same CHF experimental database used by Dominion to qualify the COBRANVRB-1 code/correlation pair. This Appendix summarizes the data evaluations that were performed to qualify the VIPRE-DNVRB-1 codekorrelation pair and to develop the corresponding DNBR design limit for the correlation.
DOM-NAF-2, APPENDIX B B-2
TABLE OF CONTENTS CLASSIFICATION/DISCLAlMER.......................................................................................................
8-2 ABSTRACT.........................................................................................................................................
8-2 TABLE OF CONTENTS......................................................................................................................
8-3 LIST OF TABLES................................................................................................................................
B-4 LIST OF FIGURES.............................................................................................................................. B-4 B.l PURPOSE..................................................................................................................................
B-6 8.2 APPLlCABlLl TY.........................................................................................................................
B-6
8.3 DESCRIPTION
OF THE WESTINGHOUSE WRB-1 CHF CORRELATION...........................
8-7
8.4 DESCRIPTION
OF VIPRE-DNVRB-1 DATABASE...................................................................
8-8 8.5 VIPRE-DNVRB-1 TEST ASSEMBLIES.....................................................................................
B-9 B.5.1 B.5.2 4x4 GEOMETRY TESTS.......................................................................................................... B-9 5x5 GEOMETRY TESTS........................................................................................................ B-10 B.6 VIPRE-D RESULTS AND COMPARISON TO COBRA.........................................................
B-11 B.7. BENCHMARK OF THE VIPRE-DNVRB-1 SUBCHANNEL MODEL......................................
8-22 B.7.1 STEADY STATE APPLICATION......................................................................................
8-22 B.7.2 MAIN STEAM LINE BREAK APPLICATION....................................................................
B-26 8.7.3 TRANSIENT APPLICATION............................................................................................. 8-26
8.8 CONCLUSION
S.......................................................................................................................
8-28 B.9 REFERENCES.........................................................................................................................
B-29 DOM.NAF.2.
APPENDIX B B-3
LIST OF TABLES Table B.5.1-1: 4x4 VIPRE-DNVRB-1 Experimental Database Table B.5.2-1: 5x5 VIPRE-DNVRB-1 Experimental Database Table B.6-1: VIPRE-DNVRB-1 M/P Ratio Results Table B.6-2: VIPRE-DNVRB-1 DNBR Design Limit Table 67.1 -1 : Range of VIPRE-D / COBRA 164 Benchmark Statepoints Table 8.8-1 : DNBR Limits for WRB-1 Table 8.8-2: Range of Validity for VIPRE-DNVRB-1 B-9 B-10 B-12 B-13 6-23 6-28 8-28 LIST OF FIGURES Figure B.6-1: Measured vs. Predicted CHF for VIPRE-DNVRB-1 Database Figure B.6-2: M/P vs. Pressure for VIPRE-DNVRB-1 Database Figure B.6-3: M/P vs. Mass Velocity for VIPRE-DNVRB-1 Database Figure 6.6-4: M/P vs. Quality for VIPRE-DNVRB-1 Database Figure B.6-5: DNBR vs. Pressure for VIPRE-DNVRB-1 Database Figure B.6-6: DNBR vs. Mass Velocity for VIPRE-DNRB-1 Database Figure 8.6-7: DNBR vs. Quality for VIPRE-DNVRB-1 Database B-15 B-16 B-17 B-18 B-19 8-20 B-21 Figure 87.1 -1 : Typical Surry VIPRE-D 19-Channel Model for Westinghouse 15x1 5 SIF Fuel Assem blies 6-24 6-25 8-27 B-27 Figure 67.1 -2: VIPRE-DNVRB-1 vs. COBRANVRB-1 for the 164 Analyzed Statepoints Figure 87.3-1 : VIPRE-D FWMAL Transient Sample Calculation Results Figure 67.3-2: VIPRE-D LOCROT Transient Sample Calculation Results DOM-NAF-2, APPENDIX B 8-4
ACRONYMS AND ABBREVIATIONS AOs CHF DNB DNBR FLC FW MAL HTRF LOCROT LOFA M/P MSLB MVG NMVG P/M PWR RWAP RWSC SIF SPS USNRC Axial Offset Envelope Critical Heat Flux Departure from Nucleate Boiling Departure from Nucleate Boiling Ratio Form Loss Coefficient Feedwater Malfunction Transient Heat Transfer Research Facility at Columbia University Locked Rotor Accident Loss of Flow Accident Ratio of Measured-to-Predicted CHF Main Steam Line Break Mixing Vane Grid Non-Mixing Vane Grid Ratio of Predicted-to-Measured CHF (equivalent to DNBR)
Pressurized Water Reactor Rod Withdrawal at Power Rod Withdrawal from Subcritical Surry Improved Fuel Surry Power Station US Nuclear Regulatory Commission DOM-NAF-2, APPENDIX B 8-5
B.l PURPOSE Dominion currently uses Westinghouse 15x1 5 OFA fuel assemblies at Surry Power Station, Units 1 and 2. This fuel product as implemented at Surry is also known as Surry Improved Fuel (SIF). The thermal-hydraulic analysis of this Westinghouse fuel product requires the use of the Westinghouse WRB-1 CHF Correlation (References B1 and 83). In fact, Westinghouse WRB-1 CHF correlation has been approved by the USNRC for use with Westinghouse 15x1 5 and 17x1 7 R grid type fuel, and with Westinghouse 14x14, 15x15 and 17x17 OFA-type fuel products (Reference 91).
To be licensed for use, a critical heat flux (CHF) correlation must be tested against experimental data that span the anticipated range of conditions over which the correlation will be applied. Furthermore, the population statistics of the database must be used to establish a departure from nucleate boiling ratio (DNBR) design limit such that the probability of avoiding departure from nucleate boiling (DNB) will be at least 95% at a 95% confidence level.
This Appendix documents Dominions qualification of the WRB-1 correlation with the VIPRE-D code.
This qualification was performed against a subset of the data from the Columbia-EPRI CHF database for Westinghouse R grid 17x1 7 and 15x1 5 fuel (Reference B2). This is the same subset of the Columbia-EPRI CHF database used by Dominion in the qualification of the WRB-1 correlation with the COBRA code (Reference 83). This Appendix summarizes the data evaluations that were performed to qualify the VIPRE-DNVRB-1 code/correlation pair, and to develop the corresponding DNBR design limits for the correlation.
B.2 APPLICABILITY Dominion intends to use the VIPRE-DNVRB-1 code/correlation pair for the analysis of Westinghouse 15x15 and 17x17 R grid type fuel, and Westinghouse 14x14, 15x15 and 17x17 OFA-type fuel products in PWR reactors. When evaluating these types of fuels outside of the range of validity of the WRB-1 CHF correlation, Dominion intends to use the VIPRE-DNV-3 code/correlation pair. W-3 is one of the CHF correlations contained in the USNRC approved generic version of VIPRE-01 (References 87 and B8).
The intended VIPRE-DNVRB-1 applications discussed in this Appendix are consistent with the generic intended applications listed in the main body of this report (Section 2.0). Also, more specifically, Dominion intends to use VIPRE-DNVRB-1 to analyze the transients delineated in Table 2.1 -1 in Section 2.0 of the main body of this report.
The qualification of the WRB-1 CHF correlation with the VIPRE-D code has been performed following the modeling guidelines described in Section 4.0 of this report. In addition, extensive code benchmark calculations have confirmed that the VIPRE-D models specified in sections 4.1 through 4.12 in the main body of this report produce essentially the same results as equivalent Dominion COBRA models. Some of these benchmarks are described in section B.7 of this Appendix.
This Appendix is submitted to the USNRC for review and approval in order to meet the USNRCs requirement #2 listed in the VIPRE-01 SER, as outlined in Section 2.2 in the main body of this report.
DOM-NAF-2, APPENDIX B B-6
8.3 DESCRIPTION
OF THE WESTINGHOUSE WRB-1 CHF CORRELATION In pressurized water reactor (PWR) cores, the energy generated inside the fuel pellets leaves the fuel rods at their surface in the form of heat flux, which is removed by the reactor coolant system flow. The normal heat transfer regime in this configuration is nucleate boiling, which is very efficient.
However, as the capacity of the coolant to accept heat from the fuel rod surface degrades, a continuous layer of steam (a film) starts to blanket the tube. This heat transfer regime, termed film boiling, is less efficient than nucleate boiling and can result in significant increases of the fuel rod temperature for the same heat flux. Since the increase in temperature may lead to the failure of the fuel rod cladding, PWRs are designed to operate in the nucleate boiling regime and protection against operation in film boiling must be provided.
The heat flux at which the steam film starts to form is called CHF or the point of DNB. For design purposes, the DNBR is used as an indicator of the margin to DNB. The DNBR is the ratio of the predicted CHF to the actual local heat flux under a given set of conditions. Thus, DNBR is a measure of the thermal margin to film boiling and its associated high temperatures. The greater the DNBR value (above 1.O), the greater the thermal margin.
The CHF cannot be predicted from first principles, so it is empirically correlated as a function of the local thermal-hydraulic conditions, the geometry, and the power distribution measured in the experiments. Since a CHF correlation is an analytical fit to experimental data, it has an associated uncertainty, which is quantified in a DNBR design limit. A calculated DNBR value greater than this design limit provides assurance that there is at least a 95% probability at the 95% confidence level that a departure from nucleate boiling will not occur.
The Westinghouse WRB-1 CHF correlation is defined in Reference B1 as:
[B.3.1]
where QCHF is the critical heat flux in Btuhr-ff, PF is a dimensionless performance factor dependent on the outer diameter of the rods and defined in Reference B1, GLoc is the local mass velocity in MlbmAf-hr, and XLoc is the local quality. The specific formulations for each one of these components, as well as the corresponding constants, are Westinghouse proprietary and can be found in Reference B1. Reference B1 discusses the application of the WRB-1 correlation form to the I-and R grid fuel assembly designs. Westinghouse WRB-1 CHF correlation has been approved by the USNRC for use with Westinghouse 15x1 5 and 17x17 Fi grid type fuel, and with Westinghouse 14xl4,15x15 and 17x17 OFA-type fuel products (Reference Bl). Its intended range of application for operating conditions is as follows (Reference 83):
1440 I Pressure I 2490 psia 0.9 I Mass Flux 53.7 Mlbm/hr-ff Local Quality 5 0.30 DOM-NAF-2, APPENDIX B 6-7
In response to concerns raised by the NRC in Reference B3, Dominion will impose two additional restrictions on the intended range of application:
0 0
VIPRE-DNVRB-1 will not be used when the local heat flux exceeds 1.O Mbtu/hr-f?.
VIPRE-DNVRB-1 will not be used for fuel with less than 1 3 mixing vane grid spacing.
The W-3 correlation is used when conditions are outside the range of the WRB-1 DNB correlation.
Specifically, the W-3 correlation is applied to the lower portion of the fuel assemblies in the rod withdrawal from subcritical event because of the bottom peaked axial power distribution assumed, and in the steam line break event because of the low pressures involved. The W-3 correlation with a correlation limit of 1.30 is used below the fuel assembly first mixing vane grid for the rod withdrawal from subcritical event. For the steam line break event, the W-3 correlation is used with a correlation limit of 1.45 in the pressure range of 500 to 1000 psia and 1.30 for pressures above 1000 psia (Reference Bll). The Westinghouse W-3 CHF correlation is described on page 10 in Reference B10.
B.4 DESCRIPTION OF VIPRE-DNVRB-1 DATABASE The W RB-1 CHF correlation was developed from a large body of rod bundle CHF data obtained at the Columbia University Heat Transfer Research Facility (HTRF) using full-scale, electrically heated rod bundle test sections (Reference 82).
The Dominion qualification of WRB-1 in VIPRE-D was performed against a subset of the data from the Columbia-EPRI CHF database for Westinghouse R grid 17x17 and 15x15 fuel (Reference 62).
Dominion analyzed 19 test series out of the 22 series used to develop the correlation; in particular, Dominion did not consider the three series of I-grid tests and as a consequence no L grid data were included in the test population. The 19 tests represent the same subset of the Columbia-EPRI CHF database used by Dominion in the qualification of the WRB-1 correlation with the COBRA code (Reference B3).
Two criteria were used to justify data deletions:
- 1) The first was consistency with the practice of the test sponsor. Certain points were excluded from the COBRWRB-1 database because they had been excluded from the THINCNVRB-1 database in Reference B1. Most excluded data were deleted under this condition. These points were also excluded from the COBFWWRB-1 database.
- 2) The second exclusion criterion was consistency of the input data in References B1 and 82.
Although some differences were expected, data points that differed by more than ten standard deviations were excluded as being probable typographical errors in Reference B2.
With the exception of the I-grid data, and the 25 data points that were thrown out under the second criterion, the VIPRE-DNVRB-1 database is the same as the one used in Reference B1 to qualify THINCNVRB-1 for R grid fuel. This is also the same database used by Westinghouse to qualify VIPRE-OlNVRB-1 (Reference 66). The same 945 statepoints used in Reference 83 by Dominion in the qualification of the WRB-1 correlation with the COBRA code were used in this calculation. Since DOM-NAF-2, APPENDIX 6 6-8
no I-grid data were included in the test population, Dominion does not intend to apply the WRB-1 correlation to Westinghouse 15x15 standard fuel.
TEST 124 125 127 131 132 133 134 140 148 153 1 46 139 B.5 VIPRE-DMIRB-1 Test Assemblies HEATED GRID NUMBER OF TESTS (inches]
(inches]
DATABASE PIN OD I TUBE LENGTH SPACING IN VIPRE-DNVRB-1 AXIAL HEAT FLUX SHAPE OD
[inches]
4 x 4 Non-Uniform 0.422 I -
96 20 32 4 x 4 Non-Uniform 0.422 I 96 20 33 36 Non-Uniform 4 x 4 Non-Uniform 0.422 I -
96 4 x 4 Non-Uniform 0.422 I -
168 26 32 4 x 4 Non-Uniform 0.422 I -
168 20 36 4 x 4 Non-Uniform 0.422 I -
168 13 35 4 x 4 Non-Uniform 0.422 I 168 32 38 4 x 4 Non-Uniform 0.422 I -
96 32 30 4 x 4 Non-Uniform 0.422 I -
168 26 70 4 x 4 Uniform 0.422 I 168 26 40 4 x 4 Non-Uniform 0.422 10.545 168 26 37 4 x 4 Non-Uniform 0.422 10.545 168 32 37 8.5.1 4x4 Geometry Tests Twelve of the nineteen tests used by Dominion to qualify the VIPRE-DNVRB-1 code/correlation pair have a 4x4 geometry. These 4x4 test bundles have essentially a 15x1 5 subchannel geometry (Reference 83, page 13). Table B.5.1-1 provides a summary of the key information about each test.
Table 6.5.1 -1 : 4x4 VIPRE-DNVRB-1 Experimental Database DOM-NAF-2, APPENDIX B 6-9
8.5.2 5x5 Geometry Tests Seven of the nineteen tests used by Dominion to qualify the VIPRE-DNVRB-1 code/correlation pair have a 5x5 geometry. These 5x5 test bundles have the same subchannel geometry as the current Westinghouse 17x1 7 R grid fuel. Table B.5.2-1 provides a summary of key information about each test.
TEST 161 156 160 157 164 162 158 Table 6.5.2-1 : 5x5 VIPRE-DNVRB-1 Experimental Database IN OD HEATED GRID NUMBER OF TESTS TUBE LENGTH SPACING IN VIPRE-DMIRB-1 AXIAL HEAT FLUX SHAPE OD
[inches]
[inches]
DATABASE
[inches]
5 x 5 Uniform 0.374 I -
168 22 71 5 x 5 Uniform 0.374 I -
168 26 70 5 x 5 Uniform 0.374 I -
96 22 65 5 x 5 Uniform 0.374 I -
96 26 76 5 x 5 Non-Uniform 0.374 I -
168 22 74 5 x 5 Non-Uniform 0.374 10.485 168 22 70 5 x 5 Uniform 0.374 10.482 96 26 63 DOM-NAFQ, APPENDIX B B-10
B.6 VIPRE-D RESULTS AND COMPARISON TO COBRA Reference 82 describes the mathematical model for each separate test section by providing the bundle and cell geometry, the rod radial peaking values, the rod axial flux shapes, the types, axial locations and form losses associated to the spacer grids, as well as the thermocouple locations.
Reference B2 provides the data for each CHF observation within a test, including power, flow, inlet temperature, pressure and CHF axial location.
Each test section was modeled for analysis with the VIPRE-D thermal-hydraulic computer code as a full assembly model following the modeling methodology discussed in Section 4 in the main body of this report. For each set of bundle data, VIPRE-D produces the local thermal-hydraulic conditions (mass velocity, thermodynamic quality, heat flux, etc.) at every axial node along the heated length of the test section. The ratio of measured-to-predicted CHF (M/P) is the variable that is normally used to evaluate the thermal-hydraulic performance of a code/correlation pair. The measured CHF is the local heat flux at a given location, while the predicted CHF is calculated by the code using the WRB-1 CHF correlation. The ratio of these two values provides the M/P ratio, which is the inverse of the DNB ratio. M/P ratios are frequently used to validate CHF correlations instead of DNB ratios, because their distribution is usually a normal distribution, which simplifies their manipulation and statistical analysis.
In addition to comparing to the experimental results, the results obtained by VIPRE-D when modeling the experiments were benchmarked against the results obtained with the COBRA code in the USNRC approved COBRA topical (Reference 83). This comparison was just a sanity check to verify that there are no suspect datapoints and that the statepoint conditions were correctly input to the code.
This section summarizes the VIPRE-D results and the associated significant statistics. In addition, this section shows a comparison to the results obtained with the COBRA code as reported in Reference 83. This section also shows the variation of the M/P ratio with each independent variable to demonstrate that there are no biases in the data. Finally, it provides the VIPRE-D overall statistics for the nineteen WRB-1 tests and generates the DNBR design limit for the WRB-1 CHF correlation with VIPRE-D.
The WRB-1 correlation was developed by Westinghouse by correlating the CHF experimental results obtained in the tests as described in Reference B1. Westinghouse also used these test data to calculate a DNBR design limit of 1.17 for the WRB-1 correlation (References B1 and B6). Dominion used a subset of this experimental data, as described in section 6.4, to develop the VIPRE-DNVRB-1 DNBR limit. Table B.6-1 summarizes the relevant statistics for each test, and calculates the aggregate statistics for the entire set of data.
One-sided tolerance theory (Reference 84) is used for the calculation of the VIPRE-DNVRB-1 DNBR design limit. This theory allows us to calculate a DNBR limit so that, for a DNBR equal to the design limit, DNB will be avoided with 95% probability at a 95% confidence level.
DOM-NAF-2, APPENDIX B B-11
Table B.6-1: VIPRE-DNVRB-1 M/P Ratio Results NUMBER OF TESTS TEST M/P RATIO M/P RATIO M/P RATIO M/P RATIO AVERAGE STDEV MAX MIN TS124 TSl25 TS127 TSl31 TSl32 32 0.9984 0.051 0 1.083 0.851 33 0.9352 0.0533 1.041 0.802 36 1.0209 0.0877 1.307 0.897 32 1.0383 0.0827 1.1 88 0.799 36 1.0382 0.1 006 1.1 85 0.804 TS161 71 1.001 2 0.0624 1.170 0.833 TS156 TSl60 Because all the statistical techniques used below assume that the original data distribution is normal, it is necessary to verify that the overall distribution for the M/P ratios is a normal distribution. To evaluate if the distribution is normal, the D normality test was applied (Reference 85). A value of D' equal to 8,160.9 was obtained for the VIPRE-DNVRB-1 database. This D' value is within the range of acceptability for 945 data points with a 95% confidence level (8,134.0 to 8,245.4)a. Thus, it is concluded that the M/P distribution for the VIPRE-DMIRB-1 database is indeed normal.
70 1.01 32 0.0780 1.163 0.81 2 65 1.0238 0.081 2 1.171 0.763 a From Table 5 in Reference B5 D Lower Limit (945) [P = 0.0251 = 7,558 + (45 / 50) x (8,198 - 7,558) = 8,134.0 D Upper Limit (945) [P = 0.9751 = 7,664 + (45 / 50) x (8,310 - 7,664) = 8,245.4 TSl57 TS164 TS162 TSl58 5x5 VIPRE-DMIRB-1 DOM-NAF-2, APPENDIX B 76 1.0222 0.0769 1.223 0.800 74 1.0468 0.0869 1.271 0.841 70 0.9825 0.071 2 1.156 0.845 63 1.01 68 0.0848 1.223 0.823 489 1.01 54 0.0794 1.271 0.763 945 1.0051 0.0827 1.307 0.760 B-12 COBRAMIRB-1 (Reference B3) 945 1.001 0 0.0838 1.287 0.745
Based on the results listed in Table B.6-1, the deterministic DNBR design limit can be calculated as:
Number of data Degrees of freedom
[B.6.1]
n 945 N
=n-1-12 932 where M/P (Jwp KN,C,p
= average measured-to-predicted CHF ratio
= standard deviation of the measured-to-predicted CHF ratios of the database
= one-sided tolerance factor based on N degrees of freedom, C confidence level, and P portion of the population protected. This number is taken from Table 1.4.4 in Reference 84.
Average WP Standard Deviation Normally, the number of degrees of freedom would be the total number of data minus one. However, because Westinghouse used these experimental data to correlate the 12 constants that appear in the WRB-1 correlation, the total number of degrees of freedom must be corrected to account for this.
In addition, the standard deviation of the database needs to be corrected accordingly to account for this reduced number of degrees of freedom:
MIP 1.005 0.083 N=n-1-12 ON= Owp- [ (n-l)/N]
Corrected Standard Deviation
[B.6.2]
0.084
= CTMP. [ (n -1) / N ]
ON Then, the DNBR design limit for the VIPRE-D and the WRB-1 correlation can be calculated as described in Table 6.6-2:
Table 8.6-2: VIPRE-DNVRB-1 DNBR Design Limit IVIPRE-DNVRB-1 1 Owens Factor IK(N,0.95,0.95) 1 1.730 I
I WRB-1 Design limit I DNBRL I=1 /(1.005-1.~30~0.084)
I 1.163 DOM-NAF-2, APPENDIX B B-13
With a large database such as this, with 945 statepoints, correcting for the number of constants in the WRB-1 correlation has no significant effect, though technically it is more conservative to make the correction. Either way, the calculated DNBR limit results in a value of 1.17.
Figures B.6-1 through B.6-4 display the performance of the M/P ratio and its distributions as a function of the pressure, mass velocity and quality. These plots show that there are no biases in the M/P ratio distribution, and that the performance of the WRB-1 CHF correlation is independent of the three variables of interest. The plots show a mostly uniform scatter of the data and no obvious trends or slopes. These plots also show that all the tests in the WRB-1 database are within 3.6 standard deviations from the average. Figures B.6-5 through 8.6-7 display the performance of the P/M ratio (i.e. the DNBR) against the major independent variables for the WRB-1 database. These plots also include a DNBR design limit line at 1.1 7. It can be seen that only 35 data points (3.70% of the database) are above the DNBR design limit, and that these data in excess of the limit are distributed over the entire range of the relevant variables.
DOM-NAF-2, APPENDIX B B-14
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B.7. BENCHMARK OF THE VIPRE-DNVRB-1 SUBCHANNEL MODEL In Section 5 of the main body of this report, the Dominion VIPRE-D models created using the selections and modeling guidelines described in Section 4 in the main body of this report provided close comparison to the Framatome ANP LYNXT code, which is a USNRC approved subchannel code. This section in Appendix B demonstrates that the Dominion VIPRE-D models created using the selections and modeling guidelines described in Section 4 in the main body of the report provide close comparison to Dominions COBRA code, which is also a USNRC approved subchannel code.
This benchmark is provided as an example to demonstrate in sufficient detail the validity of the methodology discussed in the body of this report, and it is not meant to be linked to a specific plant or fuel product.
8.7.1 STEADY STATE APPLICATION Dominion created a 19-channel model for Westinghouse 15x1 5 SIF fuel at SPS in accordance with the methodology described in Section 4 of this report. This VIPRE-D model of the 1/8h Surry core consists of 19 channels (1 5 subchannels and 4 lumped channels) and 20 rods, as shown in Figure 8.7.1-1. The axial nodalization used in this model has been customized for Westinghouse 15x15 SIF fuel assemblies and contains 73 non-uniform axial nodes with typical node lengths of 2 inches and a maximum node length of less than 6 inches. The reference axial power profile (1.55 chopped cosine) was defined by the default function provided by the VIPRE-D code.
The Westinghouse SIF fuel assembly consists of 204 fuel rods with an outside diameter of 0.422 inches arranged in a 15x15 matrix with a pin pitch of 0.563 inches. The Westinghouse SIF fuel contains several advanced design features, such as mixing vane grids (MVG). The local FLCs used in this VIPRE-D 1 9-channel model were provided by Westinghouse from full-scale hydraulic tests.
VIPRE-D benchmark calculations were performed against the Dominion COBRA code and the COBRA 19-channel model created by Dominion to model SPS cores containing Westinghouse 15x1 5 SIF fuel assemblies. This benchmark uses 164 state points obtained from the UFSAR Chapter 14 events including the reactor core safety limits, axial offset envelopes (AOs), rod withdrawal at power (RWAP), rod withdrawal from subcritical (RWSC), control rod misalignment, loss of flow accident (LOFA), and locked rotor accident (LOCROT) events to compare the performance of VIPRE-D and COBRA. These various limits and events provide sensitivity of DNB performance to the following: (a) power level (including the impact of the part-power multiplier on the allowable hot rod power FAH), pressure and temperature (reactor core safety limits); (b) axial power shapes (AOs);
(c) elevated hot rod power (misaligned rod); and (d) low flow (LOFA and LOCROT). The 164 statepoints cover the full range of conditions and axial offsets in the Surry UFSAR Chapter 14 evaluations (except for MSLB that is discussed in Section B.7.2), and were specifically selected to challenge both the WRB-1 and W-3 CHF correlations (Table 8.7.1-1).
This benchmark study showed an average deviation between VIPRE-D and COBRA of less than 0.6% in DNBR, with a maximum deviation of 3.75%. These results are well within the uncertainty typically associated with thermal-hydraulic codes, which has been quantified to be 5%
DOM-NAF-2. APPENDIX B 8-22
(Reference B9), and justify the model selections in Section 4. Figure 8.7.1-2 shows graphically the performance of VIPRE-D versus COBRA for the 164 statepoints. The close comparison of VIPRE-D to COBRA over the full range of conditions expected for UFSAR transients justifies the applications of VIPRE-D to the transients identified in Table 2.1 -1 in the main body of this report (MSLB will be discussed in Section 8.7.2).
VARIABLE Table 67.1-1 : Range of VIPRE-D / COBRA 164 Benchmark Statepoints RANGE Pressure [psia]
1800 to 2483.2 I Power [h of 2546 MWt]
I 53.4 to 144.5 I
Inlet Temperature [F]
Flow [h of Minimum Measured Flow]
505.1 to 631.7 66.8 to 100 I FAH I
1.56 to 2.1 06 I
I Axial Offset [/.I I
-76.6 to 32.2 I
DOM-NAF-2, APPENDIX B 8-23
Figure 67.1 -1. Typical Surry VIPRE-D 19-Channel Model for Westinghouse 15x15 SIF Fuel Assemblies I nspment Tube Rod 17 Channel 16 Remainder of PLB DOM-NAF-2, APPENDIX B 8-24
\\ \\ \\
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8.7.2 MAIN STEAM LINE BREAK APPLICATION The VIPRE-D 19-channel model discussed in section B.7.1 was also used to simulate the behavior of the core during a MSLB event, as it allows the modeling of the peaking and inlet boundary conditions in the fuel assemblies adjacent to the hot assembly. The two most limiting cases from a recent reload were evaluated with the VIPRE-D code, and their results compared to the COBRA results. The results obtained show a maximum deviation of 1.5% in DNBR. These results demonstrate that VIPRE-D can analyze a MSLB event, provided the model has sufficient detail surrounding the hot assembly, such as the 19-channel model described here. It is important to note that both MSLB statepoints evaluated occurred at pressures below 1000 psia, and therefore the MDNBR was evaluated with the W-3 CHF correlation, and the appropriate correlation limit was 1.45 (Reference B6).
B.7.3 TRANSIENT APPLICATION As demonstrated in Section 5.3 in the main body if this report, VIPRE-D has the capability to perform transient calculations by using boundary conditions obtained from a reactor systems code or a neutronic code. The reactor systems code provides time-dependent forcing functions for pressure, core average power, core flow rate and core inlet temperature and the neutronics code provides core power distributions and nuclear peaking factors.
VIPRE-DNVRB-1 transient capability was tested by performing two sample transient calculations.
These two transient calculations were only intended to be samples designed to exercise the transient capabilities of the VIPRE-D code and a typical VIPRE-D model created according to the guidelines discussed in Section 4 in the main body of this report. In both cases, the behavior of the VIPRE-D results was successfully compared to the behavior of the COBRA analysis of record in the UFSAR.
The first sample transient selected to perform this verification was the Feedwater Malfunction Transient (FWMAL). Forcing functions for the FWMAL transient were obtained from the SPS UFSAR. The length of the transient was 195 seconds, with a 0.5-second time step. COBRA analysis of record and VIPRE-D calculations exhibited similar behavior, and the MDNBR results show a maximum deviation of less that 0.4% (see Figure B7.3-1).
The second sample transient selected to perform this verification was the Locked Rotor Transient (LOCROT). Forcing functions for the LOCROT transient were obtained from the SPS UFSAR. The length of the transient was 9.5 seconds, with a 0.025-second time step. COBRA analysis of record and VIPRE-D calculations exhibited similar behavior, and the MDNBR results show a maximum deviation of less that 1.6% (see Figure B7.3-2).
The transient analyses demonstrate that VIPRE-DNVRB-1 is capable of performing stable transient calculations and the results obtained are essentially the same as the COBRANVRB-1 results documented in the SPS UFSAR.
DOM-NAF-2, APPENDIX B 8-26
Figure 87.3-1 : VIPRE-D FWMAL Transient Sample Calculation Results 2.8 -
2.6 -.
3 1.6 -
1.2 -
2.4 2 2 a
m s
2 s
Figure B7.3-2: VIPRE-D LOCROT Transient Sample Calculation Results 4
3.5 3
a m
z 0 I 2.5 2
1.5 0
1 2
3 4
5 6
7 8
9 10 Time [s]
DOM-NAF-2, APPENDIX B 6-27
8.8 CONCLUSION
S Dominion VIPRE-D The WRB-1 correlation has been qualified with Dominions VIPRE-D computer code. Table 8.8-1 summarizes the DNBR design limits for VIPRE-DNRB-1 that yields a 95% non-DNB probability at a 95% confidence level. The limit of 1.17 from VIPRE-D is the same limit as found with three other, approved code packages: COBRA (Reference B3), THINC (Reference Bl), and Westinghouses version of VIPRE-01 (Reference B6). Westinghouse WRB-1 CHF correlation has been approved by the USNRC for use with Westinghouse 15x15 and 17x17 R grid type fuel, and with Westinghouse 14x1 4, 15x1 5 and 17x1 7 OFA-type fuel products.
Dominion Westinghouse Westinghouse COBRA THINC VIPRE-01 Table 8.8-1 : DNBR Limits for WRB-1 DNBR limit WRB-1 CHF CORRELATION DESIGN LIMITS 1.17 1.17 1.17 1.17 Table 8.8-2 summarizes the applicability and the ranges of validity for VIPRE-DNVRB-1, which are the same as those on page 2 of the Dominion COBRA SER in Reference 63.
Table 8.8-2: Range of Validity for VIPRE-D/WRB-1 Pressure 1,440 to 2,490 Mass Velocity I
[Mlbmhr-ff]
0.9 to 3.7 Thermodynamic Quality at CHF 5 0.30 Local Heat Flux 51.0 1
[Mbtu/hr-ff]
Mixing Vane Grid Spacing [in]
> 13.0 Finally, extensive code benchmark calculations have confirmed that the VIPRE-DNVRB-1 models created using the modeling guidelines specified in Section 4 in the main body of this report produce essentially the same results as USNRC approved equivalent Dominion COBRANVRB-1 models.
DOM-NAFP, APPENDIX B B-28
B.9 B1.
- 82.
B3.
- 84.
- 85.
B6.
B7.
- 88.
B9.
B10.
B11.
REFERENCES Technical Report, W CAP-8762-P-AI New Westinghouse Correlation WRB-7 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids, F. E. Motley, et al., July 1984.
Technical Report, EPRl NP-2609, Parametric Study of CHF Data, Volume 3, Part 1; Critical Heat Flux Data, C. F. Fighetti, & D.G. Reddy, September 1982.
Topical Report, VEP-NE-3-A, Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code, R.C. Anderson & N. P. Wolfhope, July 1990.
Technical Report, Tables for Normal Tolerance Limits, Sampling Plans, and Screening, R. E.
Odeh and D. B. Owen. 1980.
Technical Report, Assessment of the Assumption of Normality (employing individual observed values), American National Standards Institute, ANSI N15.15.1974.
Technical Report, WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, Y. X. Sung, P. Schueren, and A. Meliksetian, October 1999.
Letter from C. E. Rossi (NRC) to J. A. Blaisdell (UGRA Executive Committee), Acceptance for Referencing of Licensing Topical Report, EPRl NP-2511 -CCM, VIPRE-01 : A Thermal-Hydraulic Analysis Code for Reactor Cores, Volumes 1,2, 3 and 4, May 1, 1986.
Letter from A. C. Thadani (NRC) to Y. Y. Yung (VIPRE-01 Maintenance Group), Acceptance for Referencing of the Modified Licensing Topical Report, EPRl NP-2511 -CCM, Revision 3, VIPRE-01 : A Thermal Hydraulic Analysis Code for Reactor Cores, (TAC No. M79498),
October 30, 1993.
Topical Report, VEP-NE-2-A, Statistical DNBR Evaluation Methodology, R. C. Anderson, June 1987.
Technical Report, Boiling Crisis and Critical Heat Flux, TID-25887, 1972.
Letter from A.C. Thadani (NRC) to W. J. Johnson westinghouse), Acceptance for Referencing of Licensing Topical Report, WCAP-9226-P, WCAP-9227-NP, Reactor Core Response to Excessive Secondary Steam Releases, 1989.
DOM-NAF-2, APPENDIX B 8-29