RA-15-000, Attachment 8 - DPC-NE-2005 Appendix H Robinson Plant Specific Data
| ML15075A212 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 11/30/2014 |
| From: | Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15075A221 | List: |
| References | |
| RA-15-0004 DPC-NE-2005, Appendix H | |
| Download: ML15075A212 (23) | |
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{{#Wiki_filter:Attachment 8 RA-1 5-0004 Attachment 8 DPC-NE-2005 Appendix H Robinson Plant Specific Data (Redacted) DPC-NE-2005 Duke Energy Thermal-Hydraulic Statistical Core Design Methodology APPENDIX H Robinson Plant Specific Data Advanced W 15x15 HTP Fuel Application of HTP CHF Correlation to the Advanced W 15x15 HTP Fuel Design November 2014 Note: Bracketed text, tables, and figures are "D" (Duke) and/or "A" (AREVA NP)proprietary information. This Appendix contains the plant specific data and limits for the Robinson Nuclear Plant with Advanced W 15x15 HTP fuel using the HTP critical heat flux correlation. The thermal hydraulic statistical core design analysis process was performed as described in the main body of this method report.Plant Specific Data This analysis is for the Robinson plant, a three loop Westinghouse PWR. This analysis models the 0.424 inch diameter Advanced W 15x15 HTP fuel assembly design. This assembly is a derivative of that described in the Reference H-2 AREVA Topical Report.The Advanced W HTP design incorporates the High Mechanical Performance or HMP bottom grid, M5 fuel rod cladding, and MONOBLOCTm guide tubes relative to the Reference H-2 design.Thermal Hydraulic Code and Model The VIPRE-01 thermal-hydraulic computer code described in Reference H-3 is used in this analysis. A fourteen channel model was developed for the Advanced W 15x 15 HTP fuel design comparable to the Oconee 15x15 Mark-B-HTP fuel model described in Reference H-I. Due to the fuel assembly design differences, some specific data supplementary to Reference H-I required modification. The thermal-hydraulic characteristics of the Advanced W 15x15 HTP fuel are presented in Table H-I and the model adjustments are H-I shown in Figure H-1. Table H-I includes fuel rod, control rod, and instrument guide tube diameters, the number and design of the grids, and the fuel rod length.The VIPRE-01 model approved in Reference H-I is used to analyze the Advanced W 15x15 HTP fuel with the following modifications:
- 1) The Advanced W 15x15 HTP fuel assembly geometry information and model layout as described in Table H-I and Figure H-1.2) A radial power distribution based on the Robinson fuel geometry and current peaking limits as shown in Figure H-2.3) The number of axial nodes in the model was increased due to the addition of Integral Flow Mixing (IFM) grids to the Advanced W 15x15 HTP fuel design at Robinson.4) The EPRI bulk void fraction model is used instead of the Zuber-Findlay model. The Zuber-Findlay bulk void model is not applicable for void fractions above 85%(Reference H-3). The EPRI bulk void model is essentially the same as the Zuber-Findlay bulk void model except for the equation used to calculate the drift velocity (Reference H-3). This eliminates the discontinuity at higher void fractions.
Therefore, the EPRI model provides a full range (i.e., void fraction range, 0 -1.0) of applicability required for performing DNB calculations. Also, for overall model compatibility, the EPRI subcooled void model is used. The EPRI void models are also used for the Mark-B 11, Advanced Mark-BW, Mark-B-HTP, and RFA fuel analyses as approved in Appendix D, E, F, and G of DPC-NE-2005 respectively. H-2 Critical Heat Flux Correlation The NRC approved HTP critical heat flux correlation form in Reference H-4 is used for the Advanced W 15x15 HTP analyses in VIPRE-01. This correlation was developed by AREVA for application to the HTP fuel design. The original application was to the HTP fuel designs (Advanced W 15x15 HTP and Advanced W 17x17 HTP) with the XCOBRA-IIIC code, Reference H-4. The same database was subsequently analyzed in Reference H-5 with the LYNXT thermal-hydraulic computer code. []A The HTP correlation form [ ]A,D was added to the VIPRE-01 thermal-hydraulic computer code by Duke Energy and the CHF test data base analyzed in its entirety. The results of this analysis are shown in Table H-2. The resulting VIPRE-01 average P/M is [ ],D the standard deviation [ ],D and the correlation DNBR limit is lower than the value of XCOBRA-IIIC (Reference H-4, shown on Table H-2 under XCOBRA column). Figures H-3 through H-6 graphically shows the results of this evaluation. Figure H-3 shows there is [].D Figures H-4 through H-6 show there is []D These plots also include the additional [ A, D uncorrelated data used to expand the correlation range of applicability. Similar to the XCOBRA-IIIC (Reference H-4) and LYNXT (Reference H-5) results, the VIPRE-01 results for the [ ] A, D uncorrelated data show that the correlation conservatively predicts CHF for the extended range of applicability. H-3 Based on the results shown in Table H-2 and Figures H-3 through H-6, the HTP CHF correlation can be used in DNBR calculations with VIPRE-O1 for Advanced W 15x15 HTP fuel. Table H-3 shows the correlation allowable parameter range and correlation DNBR limit with VIPRE-01. Note that the higher correlation DNBR limit will be used for non-statistical analyses of the Advanced W 15x 15 HTP fuel design with VIPRE-01.Statistical Core Design Analysis Statepoints The statepoint conditions evaluated in this analysis are listed in Table H-4. These statepoints represent the range of conditions to which the statistical DNB analyses limit will be applied. The resulting range of key parameter values analyzed is listed on Table H-7.Key Parameters and Uncertainties The key parameters and their uncertainty magnitude and associated distribution used in this analysis are listed on Table H-5. The uncertainties were selected to bound the values calculated for each parameter at the Robinson plant with the Advanced W 15x15 HTP fuel design.H-4 DNB Statistical Design Limits The statistical DNBR limit for each statepoint evaluated is listed on Table H-6. Section 1 of Table H-6 contains the 500 case runs and Section 2 contains the 5,000 case runs. All of the DNBR distributions are judged to be normally distributed. The maximum statistical DNBR value in Table H-6 for 5,000 propagations is [ ] D. Therefore, the statistical design limit using the HTP CHF correlation in VIPRE-01 for Advanced W 15x15 HTP fuel at Robinson is conservatively selected to be 1.35.H-5 REFERENCES H-1. DPC-NE-3000-PA, Revision 5a, Thermal-Hydraulic Transient Analysis Methodology, October 2012 H-2. EMF-89-164(P)(A) Revision 0, Mechanical Licensing Report for H.B. Robinson High Thermal Performance Fuel Assemblies, Advanced Nuclear Fuels Corporation, November 1989 H-3. EPRI NP-25 11 -CCM-A, VIPRE-01: A Thermal-Hydraulic Code For Reactor Cores, Vol. 1-4, Revision 4.5, February 2014 H-4 EMF-92-153(P)(A), Revision 1, HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, January 2005 H-5 BAW-10241(P)(A), Revision 1, BHTP DNB Correlation Applied with LYNXT, July 2005 H-6 FIGURE H-1 Advanced W 15x15 HTP VIPRE-01 14 Channel Model K D I o Fuel Rod a Instrument Tube ELI Flow Channel y = Channel Index x = Rod Index H-7 FIGURE H-2 Advanced W 15x15 HTP VIPRE-01 14 Channel Model Radial Power Distribution D O Fuel Rod@ Instrument Tube H-8 FIGURE H-3 VIPRE-01 Measured CHF versus Predicted CHF HTP Database H-9 FIGURE H-4 VIPRE-01 Measured to Predicted CHF versus Mass Flux HTP Database H-10 FIGURE H-5 VIPRE-01 Measured to Predicted CHF versus Pressure HTP Database H-II FIGURE H-6 VIPRE-01 Measured to Predicted CHF versus Quality HTP Database H-12 TABLE H-1 Advanced W 15x15 HTP FUEL ASSEMBLY DATA (TYPICAL)GENERAL FUEL SPECIFICATIONS Fuel rod diameter, inches (Nom.) 0.424 Thimble tube diameter, inches (Nom.) 0.544 Instrument guide tube diameter, inches (Nom.) 0.544 Fuel rod pitch, inches (Nom.) 0.563 Fuel assembly pitch, inches (Nom.) 8.466 Active Fuel Length, inches (Nom) 144.0 Fuel rod length, inches (Nom.) 151.98 Component Material Grid Ni Alloy 718 Zirc-4 Zirc-4 GENERAL FUEL CHARACTERISTICS Quantity Position 1 Lower 6 Intermediate and Upper 3 Intermediate Type HMP, Non-Mixing HTP, Castellation IFM, Castellation Fuel Rod Guide Tube Instrument Tube M5Zirc-4 Zirc-4 204 20 1 HMP = High Mechanical Performance / Structural HTP = High Thermal Performance / Structural IFM = Intermediate Flow Mixer / Non-Structural H-13 TABLE H-2 VIPRE-01 HTP Correlation Verification VIPRE-01 / XCOBRA-IIIC Statistical Results n, # of data P/M, Average predicted to measured CHF*Y (P/M)*DNBR Correlation Limit* -Statistics based on the correlated data VIPRE-01[D[D[D 1.120 XCOBRA[A[A[A 1.141 TABLE H-3 CHF Test Database Analysis Results Parameter Ranges Pressure, psia Mass Velocity, M lbm/hr-ft 2 Inlet Enthalpy, Btu/lbm Thermodynamic Quality at CHF Thermal-Hydraulic Computer Code Spacer Grid Correlation DNBR Limit 1385 to 2425 0.504 to 3.563 383.6 to 646.1 less than 0.514 VIPRE-01 Advanced W 15x15 HTP 1.141 H-14 TABLE H-4 Robinson SCD Statepoints Statepoint Number 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 Power(1)_% RTP)RCS Flow(2)% DF Core Inlet Pressure Temperature (p~sia)rF Radial Peak Axial Peak FAH D 1) 100% RTP = 2,339 Megawatts Thermal 2) 100% RCS Flow = 258,100 gpm H-15 TABLE H-5 Robinson Statistically Treated Uncertainties Parameter Reactor Power*Coolant Flow Measurement Bypass Flow Pressure Temperature Radial Peaking Measurement E F AH Axial Peaking FZ z DNBR Correlation Code/Model Uncertainty / Standard Deviation+/-0.3% / 0.18%+/-2.7% / 1.64%+/- 1.5%+/- 40 psia+/- 4 degrees F+/-4.0% / 2.43%+/- 3.0% / 1.82%+/-4.5% I 2.75%+/- 3 inches Distribution Normal Normal Uniform Uniform Uniform Normal Normal Normal Uniform Normal Normal[[A, D]D* Percentage of 1.00% RTP (7.017 MWth) wherever applied H-16 TABLE H-5 Robinson Statistically Treated Uncertainties Continued Parameter Justification Reactor Power The core power uncertainty was calculated by statistically combining various component uncertainties associated with the measurement of power such as sensor accuracy, drift, etc. Since the component uncertainties are random and are normally distributed, the combination of these uncertainties using the sum of the squares (SRSS) methodology results in a core power uncertainty that is also normally distributed. Coolant Flow Measurement Bypass Flow Pressure Temperature Radial Peaking Measurement Same approach as Core Power.The core bypass flow is the parallel flow paths in the reactor vessel non-fuel regions and is calculated based on the driving pressure drop. The core bypass uncertainty is explicitly applied in the calculation of Core Inlet flow for each state point condition. This uncertainty was conservatively applied with a uniform distribution. The pressure uncertainty was calculated by statistically combining various component uncertainties associated with the measurement of pressure such as sensor accuracy, drift, etc. The uncertainty distribution was conservatively applied as uniform.Same approach as Pressure.This uncertainty accounts for the error associated in the physics code's calculation of radial assembly and pin power as well as uncertainty from the measurement system components such as instrumentation drift or reproducibility, conversion of instrument readings to power, etc. This uncertainty is applied as a normal distribution. H-17 TABLE H-5 Robinson Statistically Treated Uncertainties Continued Parameter Justification This uncertainty accounts for the effect on peaking due to manufacturing variations in the variables affecting the heat generation rate along the flow channel and for the effect on peaking due to reduced hot channel flow area. The uncertainty is determined by statistically combining all the manufacturing tolerances. The uncertainty is normally distributed and is conservatively applied as one-sided to assure the MDNBR channel location is consistent for all cases.Axial Peaking This uncertainty accounts for the axial peak prediction uncertainty of the physics codes. The uncertainty is normally distributed. This uncertainty accounts for the possible error in interpolating on axial peak location in the maneuvering analysis. The uncertainty is one half of the physics code's axial node. The uncertainty distribution is conservatively applied as uniform.Z DNBR Correlation Code/Model This uncertainty accounts for the CHF correlation's ability to predict DNB. The uncertainty distribution is applied as normally distributed. This uncertainty accounts for the thermal-hydraulic code uncertainties and offsetting conservatisms. This uncertainty also accounts for the small DNB prediction differences between various model sizes. The uncertainty distribution is normally distributed. H-18 TABLE H-6 Robinson Statepoint Statistical Results SECTION 1 Advanced W 15x 15 HTP Fuel, HTP Critical Heat Flux Correlation 500 Case Runs Coefficient of Statistical Statepoint
- Mean G Variation DNBR I 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 H-19 TABLE H-6 Robinson Statepoint Statistical Results Continued SECTION 2 Advanced W 15x 15 HTP Fuel, HTP Critical Heat Flux Correlation 5,000 Case Runs Coefficient of Variation Statistical DNBR Statepoint
- 5 7 9 21 23 27-Mean D H-20 TABLE H-7 Parameter Core Power (%RTP)Coolant Flow (% Design)Core Exit Pressure (psia)Core Inlet Temperature
(* F)FAH, Fz, Z Robinson Key Parameter Ranges Maximum Minimum D All values listed in this table are based on the currently analyzed Statepoints. Ranges are subject to change based on future statepoint conditions. H-21}}