L-2010-113, Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix a, Safety Evaluation Report Compliance

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Turkey Point, Units 3 and 4 - License Amendment Request for Extended Power Uprate, Attachment 04; Appendix a, Safety Evaluation Report Compliance
ML103560178
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 12/14/2010
From:
Florida Power & Light Co
To:
Office of Nuclear Reactor Regulation
References
L-2010-113, L-2010-299, LAR 205
Download: ML103560178 (25)
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Turkey Point Units 3 and 4 EPU Licensing Report App. A-1 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4Turkey Point Units3 and4Extended Power UprateLicensing ReportAttachment4AppendixASafety Evaluation Report ComplianceThis coversheet plus 24pages Turkey Point Units 3 and 4 EPU Licensing Report App. A-2 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.1SAFETY EVALUATION REPORT COMPLIANCE INTRODUCTIONThis Appendix is being provided as a supplement to the codes and methods information provided in LR Section2.8.5.0 for the PTN Extended Power Uprate and is directly applicable to the PTN EPU analyses. This appendix addresses compliance with the limitations, restrictions, and conditions for the codes and methods listed below and specified in the approving safety evaluation of the applicable codes and methods (RS-001 Section2.1 Matrix8 Note7).TableA.1-1 presents an overview of the Safety Evaluation Reports (SER) by codes and methods. For each SER, the applicable report subsections and appendix subsections are listed.TableA.1-1Safety Evaluation Report Compliance SummaryNo.SubjectTopical Report(Reference)/Date of NRC AcceptanceCode(s)Limitation,Restriction,ConditionLR SectionAppendixSection Non-LOCA Thermal TransientsWCAP-7908-A (ReferenceA.1-1

)/September30,1986FACTRANYes2.8.5.4.12.8.5.4.6A.2 Non-LOCA Safety AnalysisWCAP-14882-P-A (ReferenceA.1-2

)/February11, 1999RETRANYes2.8.5.1.12.8.5.1.2 2.8.5.2.12.8.5.2.22.8.5.2.3 2.8.5.3.12.8.5.3.22.8.5.4.2A.3 Non-LOCA Safety AnalysisWCAP-7907-P-A (ReferenceA.1-3

)/July29,1983LOFTRANYes2.8.5.4.22.8.5.4.3 2.8.5.7A.4 Non-LOCA Thermal/

Hydraulics WCAP-11397-P-AReferenceA.1-14 January17,1989RTDPYes2.8.32.8.5.1.1 2.8.5.1.22.8.5.2.12.8.5.3.1 2.8.5.3.22.8.5.4.22.8.5.4.3A.8Neutron KineticsWCAP-7979-P-A (ReferenceA.1-4

)/July29,1974TWINKLENone for Non-LOCA Transient Analysis2.8.5.4.12.8.5.4.6Not Applicable Turkey Point Units 3 and 4 EPU Licensing Report App. A-3 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4ReferencesA.1-1WCAP-7908-A, "FACTRAN - A FORTRAN IV Code for Thermal Transients in a UO 2 Fuel Rod," H. G. Hargrove, December1989.1.Multi-dimensionalNeutronicsWCAP-10965-P-A (ReferenceA.1-5

)/June23,1986ANCNone for Non-LOCA Transient

Analysis2.8.5.1.22.8.5.4.3 2.8.2Not Applicable2.Non-LOCA Thermal/ HydraulicsWCAP-14565-P-A (ReferenceA.1-6

)/January19,1999VIPREYes2.8.5.1.12.8.5.1.22.8.5.3.1 2.8.5.3.22.8.5.4.12.8.5.4.3 2.8.3A.53.Steam Generator Tube RuptureWCAP-10698-P-A (ReferenceA.1-15

)/March30,1987LOFTTR2None for Steam Generator Tube Rupture2.8.5.6.2Not Applicable4.AppK SBLOCAWCAP-10079-P-A,WCAP-10054-P-A(with addenda),

WCAP-11145,WCAP-14710 (ReferencesA.1-7throughA.1-11)/May23,1985NOTRUMPYes2.8.5.6.3.3A.65.LOCA Hydraulic

ForcesWCAP-8708-P-A (ReferenceA.1-12

/June17,1977,WCAP-9735 Rev.2 (ReferenceA.1-13

)MULTIFLEX3.0Yes2.8.5.6.3.5A.76.ASTRUM BELOCAWCAP-16009-P-A (ReferenceA.1-16

)/November5,2004 WCOBRA/TRACYes2.8.5.6.3.2A.9TableA.1-1 (Continued)Safety Evaluation Report Compliance SummaryNo.SubjectTopical Report(Reference)/Date of NRC AcceptanceCode(s)Limitation,Restriction,ConditionLR SectionAppendixSection Turkey Point Units 3 and 4 EPU Licensing Report App. A-4 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.1-2WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," D. S. Huegel, et al., April1999.A.1-3WCAP-7907-P-A, "LOFTRAN Code Description," T. W. T. Burnett, et al., April1984.

A.1-4WCAP-7979-P-A, "TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," D. H. Risher, Jr. and R. F. Barry, January1975.A.1-5WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," Y. S. Liu, etal., September1986.A.1-6WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," Y. X. Sung, et al., October1999.A.1-7WCAP-10079-P-A and WCAP-10080-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," Meyer, P. E., August1985.A.1-8WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," N. Lee, et al., August1985.A.1-9WCAP-10054-P-A, Addendum2, Revision1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," C. M. Thompson, et al., July1997.A.1-10WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," S. D. Rupprecht, et al., 1986.A.1-11WCAP-14710-P-A, "1-D Heat Conduction Model for Annular Fuel Pellets," D. J. Shimeck, May1998.A.1-12WCAP-8708-P-A and WCAP-8709-A, "MULTIFLEX A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," K. Takeuchi, etal., September1977.A.1-13WCAP-9735, Rev. 2 and WCAP-9736, Rev. 1, "MULTIFLEX3.0 A FORTRAN IV Computer Program for Analyzing Thermal-Hydraulic-Structural System Dynamics Advanced Beam Model," K.Takeuchi, et al., February1998.A.1-14WCAP-11397-P-A, "Revised Thermal Design Procedure," Friedland, A. J. and Ray, S., April1999.A.1-15WCAP-10698-P-A and WCAP-10750-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," R. N. Lewis, et al., August1987.A.1-16M. E. Nissley, et. al., Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), WCAP-16009-P-A (Proprietary Version), WCAP-16009-NP-A (Non-Proprietary Version), January2005.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-5 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.2FACTRAN FOR NON-LOCA THERMAL TRANSIENTSTableA.2-1FACTRAN for Non-LOCA Thermal TransientsLimitations, Restrictions, and Conditions1."The fuel volume-averaged temperature or surface temperature can be chosen at a desired value which includes conservatisms reviewed and approved by the NRC."JustificationThe FACTRAN code was used in the analyses of the following transients for PTN: Uncontrolled Rod Withdrawal from Subcritical (PTN UFSAR Section14.1.1) and RCCA Ejection (PTN UFSAR Section14.2.6). Initial fuel temperatures used as FACTRAN input in the RCCA Ejection analysis were calculated using the NRC-approved PAD4.0 computer code, as described in WCAP-15063-P-A (ReferenceA.2-1). As indicated in WCAP-15063-P-A, the NRC has approved the method of determining uncertainties for PAD4.0 fuel temperatures.2."Table2 presents the guidelines used to select initial temperatures."JustificationIn summary, Table2 of the SER specifies that the initial fuel temperatures assumed in the FACTRAN analyses of the following transients should be "High" and include uncertainties: loss of flow, locked rotor, and rod ejection. As discussed above, fuel temperatures were used as input to the FACTRAN code in the RCCA ejection analysis for PTN. The assumed fuel temperatures, which were calculated using the PAD4.0 computer code (ReferenceA.2-1

), include uncertainties and are conservatively high. FACTRAN was not used in the loss of flow and locked rotor analyses.3."The gap heat transfer coefficient may be held at the initial constant value or can be varied as a function of time as specified in the input."JustificationThe gap heat transfer coefficients applied in the FACTRAN analyses are consistent with SER Table2. For the rod withdrawal from subcritical transient, the gap heat transfer coefficient is kept at a conservative constant value throughout the transient; a high constant value is assumed to maximize the peak heat flux (for DNB concerns) and a low constant value is assumed to maximize fuel temperatures. For the RCCA ejection transient, the initial gap heat transfer coefficient is based on the predicted initial fuel surface temperature, and is ramped rapidly to a very high value at the beginning of the transient to simulate clad collapse onto the fuel pellet.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-6 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 44."-the Bishop-Sandberg-Tong correlation is sufficiently conservative and can be used in the FACTRAN code. It should be cautioned that since these correlations are applicable for local conditions only, it is necessary to use input to the FACTRAN code which reflects the local conditions. If the input values reflecting average conditions are used, there must be sufficient conservatism in the input values to make the overall method conservative."JustificationLocal conditions related to temperature, heat flux, peaking factors and channel information were input to FACTRAN for each transient analyzed for PTN {Uncontrolled rod withdrawal from subcritical (PTN UFSAR Section14.1.1) and RCCA ejection (PTN UFSAR Section14.2.6)}.

Therefore, additional justification is not required.5."The fuel rod is divided into a number of concentric rings. The maximum number of rings used to represent the fuel is 10. Based on our audit calculations we require that the minimum of 6 should be used in the analyses."JustificationAt least 6 concentric rings were assumed in FACTRAN for each transient analyzed for PTN (Uncontrolled rod withdrawal from subcritical (PTN UFSAR Section14.1.1) and RCCA ejection (PTN UFSAR Section14.2.6.6."Although time-independent mechanical behavior (e.g., thermal expansion, elastic deformation) of the cladding are considered in FACTRAN, time-dependent mechanical behavior (e.g., plastic deformation) is not considered in the code. -for those events in which the FACTRAN code is applied (see Table1), significant time-dependent deformation of the cladding is not expected to occur due to the short duration of these events or low cladding temperatures involved (where DNBR Limits apply), or the gap heat transfer coefficient is adjusted to a high value to simulate clad collapse onto the fuel pellet."JustificationThe two transients that were analyzed with FACTRAN for PTN (Uncontrolled rod withdrawal from subcritical (PTN UFSAR Section14.1.1) and RCCA ejection (PTN UFSAR Section14.2.6)) are included in the list of transients provided in Table1 of the SER; each of these transients is of short duration. For the Uncontrolled rod withdraw al from subcritical transient, relatively low cladding temperatures are involved, and the gap heat transfer coefficient is kept constant throughout the transient. For the RCCA ejection transient, a high gap heat transfer coefficient is applied to simulate clad collapse onto the fuel pellet. The gap heat transfer coefficients applied in the FACTRAN analyses are consistent with SER Table2.TableA.2-1 (Continued)FACTRAN for Non-LOCA Thermal TransientsLimitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-7 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4ReferencesA.2-1WCAP-15063-P-A, Revision1 (with Errata) "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," J. P. Foster and S. Sidener, July2000.7."The one group diffusion theory model in the FACTRAN code slightly overestimates at beginning of life (BOL) and underestimates at end of life (EOL) the magnitude of flux depression in the fuel when compared to the LASER code predictions for the same fuel enrichment. The LASER code uses transport theory. There is a difference of about 3percent in the flux depression calculated using these two codes. When [T(centerline) - T(Surface)] is on the order of 3000°F, which can occur at the hot spot, the difference between the two codes will give an error of 100°F. When the fuel surface temperature is fixed, this will result in a 100°F lower prediction of the centerline temperature in FACTRAN. We have indicated this apparent nonconservatism to Westinghouse. In the letter NS-TMA-2026, dated January12,1979, Westinghouse proposed to incorporate the LASER-calculated power distribution shapes in FACTRAN to eliminate this non-conservatism. We find the use of the LASER-calculated power distribution in the FACTRAN code acceptable."JustificationThe condition of concern (T(centerline) - T(surface) on the order of 3000°F) is expected for transients that reach, or come close to, the fuel melt temperature. As this applies only to the RCCA ejection transient, the LASER-calculated power distributions were used in the FACTRAN analysis of the RCCA ejection transient for PTN.List of transients and accidents that use the FACTRAN program (approved in NRC SER)A.Uncontrolled RCC Assembly Bank Withdrawal from a Subcritical Condition.B.Partial Loss of Forced Reactor Coolant Flow C.Complete Loss of Forced Reactor Coolant Flow.

D.Single Reactor Coolant Pump Locked Rotor.

E.Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection)For the PTN EPU, FACTRAN was used for A and E from the approved list.TableA.2-1 (Continued)FACTRAN for Non-LOCA Thermal TransientsLimitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-8 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.3RETRAN FOR NON-LOCA SAFETY ANALYSISTableA.3-1RETRAN for Non-LOCA Safety AnalysisLimitations, Restrictions, and Conditions1."The transients and accidents that Westinghouse proposes to analyze with RETRAN are listed in this SER (Table1) and the NRC staff review of RETRAN usage by Westinghouse was limited to this set. Use of the code for other analytical purposes will require additional justification."JustificationThe transients listed in Table1 of the SER are:*Feedwater system malfunctions

  • Excessive increase in steam flow
  • Steam line break
  • Loss of external load/turbine trip*Loss of offsite power*Loss of normal feedwater flow
  • Feedwater line rupture
  • Control rod cluster withdrawal at power
  • Inadvertent increase in coolant inventory
  • Inadvertent opening of a pressurizer relief or safety valve
  • Steam generator tube rupture The transients analyzed for PTN using RETRAN are:
  • Feedwater system malfunctions
  • Excessive increase in steam flow
  • Steam line break
  • Loss of external electrical load/Turbine trip
  • Loss of all alternating current power to the station auxiliaries
  • Locked rotor accident
  • Uncontrolled rod withdrawal at power As each transient analyzed for PTN using RETRAN matches one of the transients listed in Table1 of the SER, additional justification is not required.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-9 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4ReferencesA.3-1WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," S. L. Davidson (Ed.), July1985.2."WCAP-14882 describes modeling of Westinghouse designed 4-, 3, and 2-loop plants of the type that are currently operating. Use of the code to analyze other designs, including the Westinghouse AP600, will require additional justification."JustificationThe PTN consists of two 3-loop Westinghouse-designed units that were "currently operating" at the time the SER was written (February11,1999). Therefore, additional justification is not required.3."Conservative safety analyses using RETRAN are dependent on the selection of conservative input. Acceptable methodology for developing plant-specific input is discussed in WCAP-14882 and in Reference14 [WCAP-9272-P-A]. Licensing applications using RETRAN should include the source of and justification for the input data used in the analysis."JustificationThe input data used in the RETRAN analyses performed by Westinghouse came from both PTN and Westinghouse sources. Assurance that the RETRAN input data is conservative for PTN is provided via Westinghouse's use of transient-specific analysis guidance documents. Each analysis guidance document provides a description of the subject transient, a discussion of the plant protection systems that are expected to function, a list of the applicable event acceptance criteria, a list of the analysis input assumptions (e.g., directions of conservatism for initial condition values), a detailed description of the transient model development method, and a discussion of the expected transient analysis results. Based on the analysis guidance documents, conservative plant-specific input values were requested and collected from the responsible PTN and Westinghouse sources. Consistent with the Westinghouse Reload Evaluation Methodology described in WCAP-9272-P-A (ReferenceA.3-1), the safety analysis input values used in the PTN analyses were selected to conservatively bound the values expected in subsequent operating cycles.TableA.3-1 (Continued)RETRAN for Non-LOCA Safety AnalysisLimitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-10 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.4LOFTRAN FOR NON-LO CA SAFETY ANALYSISTableA.4-1LOFTRAN for Non-LOCA Safety AnalysisLimitations, Restrictions, and Conditions1."LOFTRAN is used to simulate plant response to many of the postulated events reported in Chapter15 of PSARs and FSARs, to simulate anticipated transients without scram, for equipment sizing studies, and to define mass/energy releases

for containment pressure analysis. The Chapter15 events analyzed with LOFTRAN are:*Feedwater System Malfunction

  • Excessive Increase in Steam Flow
  • Steamline Break
  • Loss of External Load
  • Loss of Offsite Power*Loss of Normal Feedwater*Feedwater Line Rupture
  • Locked Pump Rotor
  • Rod Withdrawal at Power
  • Rod Drop
  • Startup of an Inactive Pump
  • Inadvertent ECCS Actuation
  • Inadvertent Opening of a Pressurizer Relief or Safety ValveThis review is limited to the use of LOFTRAN for the licensee safety analyses of the Chapter15 events listed above, and for a steam generator tube rupture-"ComplianceFor PTN, the LOFTRAN code was used in the analyses of the rod cluster control assembly drop transient, ATWS (loss of external load and loss of normal feedwater), and the uncontrolled rod withdrawal at power primary overpressurization case. As each of these transients matches one of the transients listed in the SER, additional justification is not required.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-11 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.5VIPRE FOR NON-LOCA THERMAL/HYDRAULICSTableA.5-1VIPRE for Non-LOCA Thermal/HydraulicsLimitations, Restrictions and Conditions1."Selection of the appropriate CHF correlation, DNBR limit, engineered hot channel factors for enthalpy rise and other fuel-dependent parameters for a specific plant application should be justified with each submittal."ComplianceThe WRB-1 correlation with a 95/95 correlation limit of 1.17 was used in the DNB analyses for the Turkey Point 15x15 DRFA and 15x15 Upgrade fuel. The use of the WRB-1 DNB correlation was approved in WCAP-8762-P-A (ReferenceA.5-2). Applicability of WRB-1 to Upgrade fuel was established through the Fuel Criterion Evaluation Process (FCEP) in LTR-NRC-04-8 (ReferenceA.5-3). For conditions where WRB-1 is not applicable, analyses were performed using approved secondary CHF correlations (such as ABB-NV and WLOP) in compliance with the SER conditions licensed for use in the VIPRE code. (WCAP-14565-P-A and its Addendum2-P-A, ReferenceA.5-4

).The use of the plant specific hot channel factors and other fuel dependent parameters in the DNB analysis for the Turkey Point 15x15 fuel were justified using the same methodologies as for previously approved safety evaluations of other Westinghouse three-loop plants using the same fuel design.2."Reactor core boundary conditions determined using other computer codes are generally input into VIPRE for reactor transient analyses. These inputs include core inlet coolant flow and enthalpy, core average power, power shape and nuclear peaking factors. These inputs should be justified as conservative for each use of VIPRE."JustificationThe core boundary conditions for the VIPRE calculations for the 15x15 fuel are all generated from NRC-approved codes and analysis methodologies. Conservative reactor core boundary conditions were justified for use as input to VIPRE. Continued applicability of the input assumptions is verified on a cycle-by-cycle basis using the Westinghouse reload methodology described in WCAP-9272-P-A (ReferenceA.5-1

).3."The NRC Staff's generic SER for VIPRE set requirements for use of new CHF correlations with VIPRE. Westinghouse has met these requirements for using WRB-1, WRB-2 and WRB-2M correlations. The DNBR limit for WRB-1 and WRB-2 is 1.17. The WRB-2M correlation has a DNBR limit of 1.14. Use of other CHF correlations not currently included in VIPRE will require additional justification."JustificationAs discussed in response to Condition 1, the WRB-1 correlation with a limit of 1.17 was used as the primary correlation in the DNB analyses of 15x15 DRFA and Upgrade fuel for Turkey Point. For conditions where WRB-1 is not applicable, analyses were performed using approved secondary CHF correlations licensed for the VIPRE code in ReferenceA.5-4

.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-12 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4ReferencesA.5-1WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," S. L. Davidson (Ed.), July1985.A.5-2WCAP-8762-P-A, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," F. E. Motley et. al., July1984.A.5-3LTR-NRC-04-8, "Fuel Criterion Evaluation Process (FCEP) Notification of the 15x15 Upgrade Design (Proprietary/Non-Proprietary)," James A. Gresham, February6,2004.A.5-4WCAP-14565-P-A Addendum 2-P-A, "Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications," A. Leidich, et. al., April2008.4."Westinghouse proposes to use the VIPRE code to evaluate fuel performance following postulated design-basis accidents, including beyond-CHF heat transfer conditions. These evaluations are necessary to evaluate the extent of core damage and to ensure that the core maintains a coolable geometry in the evaluation of certain accident scenarios. The NRC Staff's generic review of VIPRE did not extend to post CHF calculations. VIPRE does not model the time-dependent physical changes that may occur within the fuel rods at elevated temperatures. Westinghouse proposes to use conservative input in order to account for these effects. The NRC Staff requires that appropriate justification be submitted with each usage of VIPRE in the post-CHF region to ensure that conservative results are obtained."JustificationFor application to Turkey Point safety analysis, the use of VIPRE in the post-critical heat flux region is limited to the peak clad temperature calculation for the locked rotor transient. The calculation demonstrated that the peak clad temperature in the reactor core is well below the allowable limit to prevent clad embrittlement. VIPRE modeling of the fuel rod is consistent with the model described in WCAP-14565-P-A and included the following conservative

assumptions:*DNB was assumed to occur at the beginning of the transient,

  • Film boiling was calculated using the Bishop-Sandberg-Tong correlation,
  • The Baker-Just correlation accounted for heat generation in fuel cladding due to zirconium-water reaction.Conservative results were further ensured with the following input:
  • Fuel rod input based on the maximum fuel temperature at the given power,
  • The hot spot power factor was equal to or greater than the design linear heat rate,
  • Uncertainties were applied to the initial operating conditions in the limiting direction.TableA.5-1 (Continued)VIPRE for Non-LOCA Thermal/HydraulicsLimitations, Restrictions and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-13 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.6NOTRUMP FOR SMALL BREAK LOCANOTRUMP SER Restriction Compliance SummaryThe following table contains a synopsis of the NRC imposed Safety Evaluation Report (SER) restrictions/requirements and the Westinghouse compliance status related to these issues. Not all the items identified are clearly SER restrictions, but sometimes state the NRC's interpretation of the Westinghouse Evaluation Methodology utilized for a particular aspect of the Small Break Loss Of Coolant Accident (LOCA) Evaluation Model.TableA.6-1WCAP-10054-P-A and WCAP-10079-P-A (ReferencesA.6-1 and A.6-2)Limitations, Restrictions, and ConditionsWCAP-10054-P-A is titled "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," and is dated August1985. The following summarizes the SER restrictions and requirements associated with this WCAP:1.SER Wording (Page6)"The use of a single momentum equation implies that the inertias of the separate phases can not be treated. The model therefore would not be appropriate for situations when separate inertial effects are significant. For the small break transients, these effects are not significant."SER Compliance Inherent compliance due to the use of a single momentum equation.2.SER Wording (Page8)"To assure the validity of this application, the bubble diameter should be on the order of 10-1-2cm. As long as steam generator tube uncovery (concurrent with a severe depressurization rate) does not occur, this option is acceptable."SER Compliance Westinghouse complies with this restriction for all AppendixK licensing basis calculations. Typical AppendixK calculations do not undergo a significant secondary side system depressurization in conjunction with steam generator tube uncovery due to the modeling methodology utilized.3.SER Wording (Page14)"The two phase multiplier used is the Thom modification of the Martinelli-Nelson correlation. This model is acceptable per 10CFR50 AppendixK for LOCA analysis at pressure above 250psia."SER Compliance The original NOTRUMP model was limited to no less than 250psia since the model, as contained in the NOTRUMP code, did not contain information below this range. Westinghouse extended the model to below 250psia, as allowed by AppendixK paragraphI-C-2, and reported these modifications to the NRC via the 1995 annual reporting period (NSD-NRC-96-4639).

Turkey Point Units 3 and 4 EPU Licensing Report App. A-14 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 44.SER Wording (Page16)"Westinghouse, however, has stated that the separator models are not used in their SBLOCA analyses."SER Compliance Westinghouse does not model the separators in the secondary side of the steam generators for AppendixK Small Break LOCA analyses; therefore, compliance exists.5.SER Wording (Pages16-17)"Axial heat conduction is not modeled." and "Deletion of clad axial heat conduction maximizes the peak clad temperature."SER Compliance The Westinghouse Small Break LOCA is comprised of two computer codes, the NOTRUMP code which performs the detailed system wide thermal hydraulic calculations and the LOCTA code which performs the detailed fuel rod heatup calculations. The NOTRUMP code does not model axial conduction in the fuel rod and therefore complies. The LOCTA code has always accounted for axial conduction as is clearly stated in WCAP-14710-P-A which supplements the original NOTRUMP documentation.6.SER Wording (Page17) "...; critical heat flux, W-2, W-3, or Macbeth, or GE transient CHF (the W-2 and W-3 correlations are used for licensing evaluations);..."SER Compliance

The information presented her e indicates that the NRC apparently misstated that Westinghouse was utilizing the W-2,W-3 correlations for Critical Heat Flux (CHF) in the fuel rod heat transfer model. A review of the analyses performed by Westinghouse, including those in WCAP-11145-P-A, indicates that the Macbeth CHF correlation has been utilized for all AppendixK analyses performed by Westinghouse. This is consistent with the slab heat transfer map as described in WCAP-10054-P-A. In addition, the Macbeth correlation is specifically called out in AppendixK I-C-4-4 as an acceptable CHF model.In a supplemental response to NRC questions (Specifically question440.1 found in AppendixA of WCAP-10054-P-A, PageA-10), a description of the core model describes the Macbeth as being utilized as the CHF correlation in the NOTRUMP Small Break LOCA model.7.SER Wording (Page21)"The standard continuous contact model is not appropriate for vertical flow,..."

SER Compliance The standard continuous contact flow links are not utilized when modeling vertical flow in the AppendixK NOTRUMP Evaluation Model analyses; therefore, compliance is demonstrated.TableA.6-1 (Continued)WCAP-10054-P-A and WCAP-10079-P-A (ReferencesA.6-1 and A.6-2)Limitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-15 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 48.SER Wording (Page27)"..., the hardwired choice of one fuel pin time step per coolant time step should result in sufficient accuracy."SER Compliance The NOTRUMP code continues to utilize only one fuel pin time step per coolant time step and therefore complies with this requirement.9.SER Wording (Page47)"The code options available to the user but not applied in licensing evaluations were not reviewed."SER Compliance Westinghouse complies with this requirement.

10.SER Wording (Page53)

"4. Steam Interaction with ECCS Water, a. Zero Steam Flow in the Intact Loops While Accumulators Discharge Water."SER Compliance Per paragraphI-D-4 AppendixK, the following is stated:"During refill and reflood, the calculated steam flow in unbroken reactor coolant pipes shall be taken to be zero during the time that accumulators are discharging water into those pipes unless experimental evidence is available regarding the realistic thermal-hydraulic interaction between the steam and the liquid. In this case, the experimental data may be used to support an alternate assumption."As can be seen, the specific AppendixK wording can be considered applicable to Large Break LOCAs only since Small Break LOCAs do not undergo a true refill/reflood period. However, the Westinghouse Small Break LOCA Evaluation Model methodology is such that for break sizes in which the intact loop seal restriction is not removed (WCAP-11145-P-A Page2-11), steam flow through the intact loop(s) is automatically (artificially) restricted via the loop seal model. While not specifically limited to zero, the flow is drastically reduced via the application of the artificial loop seal restriction model.For breaks sizes above which the loop seal restriction is removed (typically 6inch diameter breaks), this criterion is not explicitly adhered to. The implementation of the COSI condensation model into NOTRUMP (As approved by the NRC in WCAP-10054-P-A, Addendum2, Revision1), which is based on additional experimental documentation and improved modeling techniques, more accurately models the interaction of steam with Emergency Core Cooling Water in the cold leg region. This experimental documentation supports the more accurate modeling of steam/water interaction in the cold leg region as allowed by AppendixK. Note however that even with the COSI condensation model active, the accumulator injection condensation model still utilizes the conservative model as originally licensed in the NOTRUMP code.TableA.6-1 (Continued)WCAP-10054-P-A and WCAP-10079-P-A (ReferencesA.6-1 and A.6-2)Limitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-16 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 411.SER Wording (Page7 of enclosure2)"Per generic letter83-35, compliance with Action ItemII.K.3.31 may be submitted generically. We require that the generic submittal include validation that the limiting break location has not shifted away from the cold legs to the hot or pump suction legs."SER Compliance Westinghouse submitted WCAP-11145-P-A in support of generic letter83-35 Action ItemII.K.3.31. As part of this effort, verification was provided which documented that the cold leg break location remains limiting.WCAP-10054-P-A, Addendum2, Revision1 (ReferenceA.6-3

)WCAP- 10054-P-A, Addendum2, Revision1 is titled "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," and is dated July1997. The following summarizes the SER restrictions and requirements associated with this WCAP:1.SER Wording (Page3)"It is stated in Ref.5 that the range of injection jet velocities used in the experiments brackets the corresponding rates in small break LOCAs for Westinghouse plants and that the model will be used within the experimental range. Also in References1 and5 Westinghouse submitted analyses demonstrating that the condensation efficiency is virtually independent of RCS pressure and state that the COSI model will be applied within the pressure range of 550 to 1200psia."SER Compliance The coding implementation of the COSI model correlation in the NOTRUMP model restricts the application of the COSI condensation model to a default pressure range of 550 to 1200psia and limits the injection flow rate to a default value of 40lbm/sec-loop. The value of 40lbm/sec-loop corresponds to the 30ft./sec velocity utilized in the COSI experiments. As such, the default NOTRUMP implementation of the COSI condensation model complies with the applicable SER restrictions.WCAP-11145-P-A (ReferenceA.6-4

)WCAP-11145-P-A, is titled "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With The NOTRUMP Code," and is dated 1986. No specific SER restrictions were provided by the NRC as part of this WCAP review; however, the SER contains verification that the requirements of ItemII.K.3.31 have been satisfied (i.e. break location study).TableA.6-1 (Continued)WCAP-10054-P-A and WCAP-10079-P-A (ReferencesA.6-1 and A.6-2)Limitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-17 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4ReferencesA.6-1WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," N. Lee, et al., August1985.A.6-2WCAP-10079-P-A and WCAP-10080-A, "NOTRUMP - A Nodal Transient Small Break And General Network Code," Meyer, P. E., August1985.A.6-3WCAP-10054-P-A, Addendum2, Revision1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," C. M. Thompson, et al., July1997.A.6-4WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," S. D. Rupprecht, et al., 1986.A.6-5WCAP-14710-P-A, "1-D Heat Conduction Model for Annular Fuel Pellets," D. J. Shimeck, May1998.1.SER Wording (Page5) "We therefore, find that the requirements of NUREG-0737, ItemI.K3.31, as clarified by Generic Letter83-35, have been satisfied."SER Compliance "We find that a condition of the safety evaluation for NOTRUMP as applied to ItemII.K.3.30 has been satisfied. The limiting cold leg break size for a 4-loop plant was reanalyzed at pump suction and at hot leg locations. The results confirmed that the cold leg break was limiting."WCAP-14710-P-A (ReferenceA.6-4

)WCAP-14710-P-A, is titled "1-D Heat Conduction Model for Annular Fuel Pellets," and is dated May1998. No specific SER restrictions are provided by the NRC in this document; however, a conclusion was reached regarding the modeling of annular pellets during Small Break LOCA event.1.SER Wording"Based on its conclusions that the explicit modeling of annular pellets, as described in WCAP -4710(P), provides a more realistic representation in W AppendixK ECCS evaluation models of the annular pellets, while retaining conservatism in those evaluation models, the staff finds that the explicit modeling of annular pellets, as described in WCAP-14710(P), in W AppendixK LOCA evaluation models permits those models to continue to satisfy the regulations to which they were approved, and is, therefore, acceptable for incorporation into those models."SER Compliance Westinghouse performs sensitivity studies to assess the impact of modeling annular pellets on plant specific analyses.TableA.6-1 (Continued)WCAP-10054-P-A and WCAP-10079-P-A (ReferencesA.6-1 and A.6-2)Limitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-18 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.7MULTIFLEX FOR LOCA HYDRAULIC FORCESThe NRC Safety Evaluation Report (SER) for the MULTIFLEX1.0 Evaluation Model can be found in the front of WCAP-8708 Rev.2 (ReferenceA.7-1). This SER stipulates a number of conditions and limitations on the use of the MULTIFLEX1.0 Evaluation Model for licensing basis calculations. The following is a review of these SER restrictions and requirements.TableA.7-2MULTIFLEX1.0Limitations, Restrictions and Conditions 1.SER Restriction - Use of Corrected Sonic Velocity (SER, page11)SER Wording - "The sonic velocity, or wave speed, computed with the empirical equation of state was not consistent with the 1967 ASME Steam Tables. The corrected sonic velocity data is required for a licensing calculation."SER Compliance - The MULTIFLEX code has been changed (prior to the issuance of Revision1 to WCAP-8708) to compute revised sonic velocity. Therefore, Westinghouse is in compliance with this restriction.

2.SER Restriction - Lower Plenum Modeling (SER, page12)SER Wording - "In the modeling region from the downcomer annulus to the lower plenum, the equivalent pipe network provided an artificially short transport distance across the length of the lower plenum. The correct radial transport distance, the diameter of the pressure vessel, is required in the model for a licensing calculation."SER Compliance - Westinghouse does not use the "artificially short" lower plenum length cited in the SER. Therefore, it can be concluded that Westinghouse is in compliance with this modeling requirement.

3.SER Restriction - 10Mass Point Downcomer (SER, page12, 18, 19)SER Wording - "The peak lateral force for a calculation using a 10mass point representation for the core support barrel shows an increase in loading of 4% over the reference5 mass point case. The NRC, therefore, requires a 10mass point model be used for a coupled licensing calculation."SER Compliance - Standard methodology uses a 10mass point structural model. Therefore, Westinghouse is in compliance with this requirement.

4.SER Restriction - 1Millisecond Break Opening Time (BOT) (SER, page13)SER Wording - "The use of a one millisecond opening time, as specified by Westinghouse, is required for a licensing calculation. Longer break opening times will not be considered unless Westinghouse demonstrated that the proposed break opening time with current equivalent pipe network adequately predicts the results of applicable experimental data."SER Compliance - Standard methodology uses a 1millisecond BOT. Therefore, Westinghouse is in compliance with this restriction.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-19 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4MULTIFLEX3.0 ApplicationsAs indicated in the SER of WCAP-15029-P-A (ReferenceA.7-3), the WCAP-9735, Rev.2 (ReferenceA.7-2) topical was submitted for NRC review and subsequently withdrawn. As stated in the SER, "Evaluation of the MULTIFLEX3.0 methodology is not a requisite for concluding that WCAP-15029 is acceptable." The Staff's discussion of MULTIFLEX3.0 is shown below:"The MULTIFLEX3.0 program is described as a more sophisticated analysis tool for LOCA hydraulic force calculations than the currently approved version, MULTIFLEX1.0. WCAP-15029 indicates that the MULTIFLEX3.0 program enhancements of MULTIFLEX1.0 include: the use of a two dimensional flow network to represent the vessel downcomer region in lieu of a collection of one dimensional parallel pipes; the allowance for non-linear boundary conditions at the vessel and downcomer interface at the radial keys and the upper core barrel flange in lieu of simplified linear boundary conditions; and the allowance for vessel motion in lieu of rigid vessel assumptions. WCAP-15029 indicates that these modifications are included in the MULTIFLEX3.0 program that is used to estimate the LOCA hydraulic forces on the vessel and consequential forces induced on the fuel and reactor vessel internal structures. The staff concurs with the WOG that MULTIFLEX3.0 provides a more accurate and realistic modeling approach.

On this basis, and considering that MULTIFLEX3.0 is based on the previously approved MULTIFLEX1.0, the staff considers the application of MULTIFLEX3.0 with the WCAP-15029 methodology reasonable and acceptable."Only one of the four SER restrictions in WCAP-15029-P-A (ReferenceA.7-3) applies to analyses performed using MULTIFLEX3.0. Limitation number2 reads: "The noding to be used in the representation of the loading is demonstrated to be adequate by performing nodalization sensitivity studies or by some other acceptable methodology."The current nodalization employed in the Westinghouse baffle-former bolting analyses has been validated through a series of calculations. Westinghouse has verified that the current MULTIFLEX code version produces equivalent results to those used in the original development of MULTIFLEX3.0 modeling features, despite several changes in operating system and computer platform. Westinghouse has demonstrated that the current standard nodalization 5.SER Restriction - Use of "Question18" Input Parameters (SER, page12). Question18 establishes a line-by-line review of MULTIFLEX input. Parameters, identifying those that are "Required for design basis blowdown analysis"SER Wording - "The response to Question18 of reference4 is to be included in the MULTIFLEX report to identify the acceptable input option for a licensing calculation."SER Compliance - The inputs used in the response to Question18 were reviewed against the MULTIFLEX inputs established as Westinghouse's current methodology. We can state that our current models conservatively bound the requirements for licensing basis calculations as described in the MULTIFLEX SER. Therefore, Westinghouse is in compliance with this restriction.TableA.7-2 (Continued)MULTIFLEX1.0Limitations, Restrictions and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-20 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4produces equivalent results to those used in original test cases. Westinghouse has performed a series of sensitivity studies on MULTIFLEX3.0 models using the current nodalization. Also, the historical model validation cases were found to yield conservative results relative to test data. This collection of documentation supports the conclusion that analyses performed to the current nodalization meet the limitation in WCAP-15029-P-A (ReferenceA.7-3

).MULTIFLEX3.0 has also been accepted for use in other applications which are limited by the same acceptance criteria, i.e. fuel qualification. The Control Rod Insertion program, documented in WCAP-15245 (ReferenceA.7-4), was performed using MULTIFLEX3.0 and the analyses were reviewed and accepted by the Staff (ReferenceA.7-5). These analyses have been used as a template for additional applications limited by the same acceptance criteria.The use of break opening times greater than 1millisecond has also been approved by the US-NRC (ReferenceA.7-6) for baffle barrel-bolting analyses. However, the use of longer break opening times is not approved for use on a generic basis. Such applications will require additional justification.References A.7-1WCAP-8708-P-A and WCAP-8709-A, "MULTIFLEX A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," K. Takeuchi, etal., September1977.A.7-2WCAP-9735, Rev.2 and WCAP-9736, Rev.1, "MULTIFLEX3.0 A FORTRANIV Computer Program for Analyzing Thermal-Hydraulic-Structural System Dynamics Advanced Beam Model," K.Takeuchi, et al., February1998.A.7-3WCAP-15029-P-A, WCAP-15030-NP-A, "Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions," December1998.A.7-4WCAP-15245 (Proprietary), WCAP-15246 (Non-proprietary), "Control Rod Insertion Following a Cold Leg LBLOCA, D. C. Cook, Units1 and2," May28,1999.A.7-5Letter from John F. Stang (US-NRC) to Robert P. Powers (Indiana Michigan Power Company), "Issuance of Amendments - Donald C. Cook Nuclear Plant, Units1 and2 (TAC Nos.MA6473 andMA6474)," December23,1999.A.7-6WCAP-14748-P-A, Revision0, WCAP-14749-NP-A, Revision0, "Justification for Increasing Postulated Break Opening Times in Westinghouse Pressurized Water Reactors," December1998.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-21 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.8REVISED THERMAL DESIGN PROCEDURE FOR NON-LOCA THERMAL HYDRAULICSTableA.8-1Revised Thermal Design Procedure for Non-LOCA Thermal HydraulicsLimitations, Restrictions, and Conditions1."Sensitivity factors for a particular plant and their ranges of applicability should be included in the Safety Analysis Report or reload submittal.JustificationSensitivity factors were evaluated using the WRB-1 and ABB-NV correlations and the VIPRE code for parameter values applicable to the 15x15 DFRA and UPGRADE fuel at EPU conditions. These sensitivity factors were used to determine the RTDP design limit DNBR values. The RTDP design limit DNBR values will be included in the Turkey Point FSAR.2."Any changes in DNB correlation, THINC-IV correlations, or parameter values listed in Table3-1 of WCAP-11397 outside of previously demonstrated acceptable ranges require re-evaluation of the sensitivity factors and of the use of Equation(2-3) of the topical report."JustificationBecause the VIPRE code was used to replace the THINC-IV code, sensitivity factors were evaluated for using the VIPRE code. VIPRE has been demonstrated to be equivalent to the THINC-IV code in WCAP-14565-P-A (ReferenceA.8-1). See the response to condition 3 for a discussion of the use of Equation(2-3) of the topical report. Evaluations using both WRB-1 and ABB-NV correlations were done in comp liance with WCAP

-11397 methodology.3."If the sensitivity factors are changed as a result of correlation changes or changes in the application or use of the THINC code, then the use of an uncertainty allowance for application of Equation(2-3) must be re-evaluated and the linearity assumption made to obtain Equation(2-17) of the topical report must be validated.JustificationEquation(2-3) of WCAP-11397-P-A (ReferenceA.8-2) and the linearity approximation made to obtain Equation(2-17) were confirmed to be valid for the Turkey Point EPU using the combination of the VIPRE code and the WRB-1 correlation, as well as the ABB-NV correlation. 4."Variances and distributions for input parameters must be justified on a plant-by-plant basis until generic approval is obtained."JustificationThe plant specific variances and distributions were justified for the EPU and are presented in Section2.8.3.5."Nominal initial condition assumptions apply only to DNBR analyses using RTDP. Other analyses, such as overpressure calculations, require the appropriate conservative initial condition assumptions."JustificationNominal conditions were only applied to the DNBR analyses which used RTDP.

Turkey Point Units 3 and 4 EPU Licensing Report App. A-22 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4ReferencesA.8-1WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic safety Analysis," Y. X. Sung, et al., October1999.A.8-2WCAP-11397-P-A, "Revised Thermal Design Procedure," Friedland, A. J. and Ray, S., April1999.6."Nominal conditions chosen for use in analyses should bound all permitted methods of plant operation.JustificationBounding nominal conditions were used in the DNBR analyses using RTDP, consistent with the proposed methods of plant operation for the EPU.7."The code uncertainties specified in Table3-1 (of WCAP-11397-P) (+/- 4percent for THINC-IV and +/- 1percent for transients) must be included in the DNBR analyses using RTDP."JustificationThe code uncertainties specified in Table3-1 of WCAP-11397-P-A (ReferenceA.8-2) remained unchanged and were included in the DNBR analyses using RTDP. The THINC-IV uncertainty was applied to VIPRE, based on the equivalence of the VIPRE model approved in WCAP-14565-P-A to THINC-IV.TableA.8-1 (Continued)Revised Thermal Design Procedure for Non-LOCA Thermal HydraulicsLimitations, Restrictions, and Conditions Turkey Point Units 3 and 4 EPU Licensing Report App. A-23 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4A.9BEST-ESTIMATE LARGE BREAK LOCAThe following discussion of the applicability limits and usage conditions imposed on the ASTRUM methodology used for the Large Break LOCA analysis is fashioned after the discussion in Section13-3 of the ASTRUM topical (WCAP-16009-P-A). Only those limits and conditions which have been determined as applicable to the ASTRUM methodology (discussed in Section13-3 of WCAP-16009-P-A and approved by the NRC in Section4.0 of the ASTRUM SER) are addressed below.TableA.9-1Best-Estimate Large Break LOCA - Applicability Limits1."The use of the WCOBRA/TRAC EM for long term cooling licensing analyses is not covered in this review."The WCOBRA/TRAC EM was used for the Large Break LOCA licensing. The WCOBRA/TRAC thermal-hydraulic computer code was used in the post-LOCA analyses for analyzing the switch to cold leg recirculation. The approach used in the post-LOCA analyses for analyzing the switch to cold leg recirculation is consistent with the method used for the Prairie Island fuel transition: U.S. Nuclear Regulatory Commission, "Prairie Island Nuclear Generating Plant, Units1 and2 - Issuance of AMENDMENTS RE: Technical Specifications Changes to Allow Use of Westinghouse 0.422-inch OD 14X14 VANTAGE+ Fuel (TAC Nos.MD9142 andMD9143)," July1,2009 (ML091460809). As such, this limit is met.2."Our review did not cover the use of the WCOBRA/TRAC EM for small break LOCA licensing analyses."The WCOBRA/TRAC EM was used for the Large Break LOCA licensing analysis, but not the small break LOCA analysis. As such, this applicability limit is met.3."Section2.4.4 of this SER [for WCAP-14449-P-A] discusses that ranges and biases of parameters were based on data, including UPTF and CCTF data. Of particular concern is the ranging of interfacial drag and condensation, which is based on UPTF and CCTF data. In a letter dated April8,1999, to assure that the 2-loop version of the methodology would not be applied for heat generation rates higher than covered by the UPTF and CCTF data, W proposed to limit the application of the UPI methodology to nominal power levels of 1980MWt, low power region average heat generation rate of less than 6.9kW/ft, and maximum analyzed linear heat generation rates of 17kW/ft. We find the proposed limits are acceptable because they are consistent with the range of the UPTF and CCTF data. We also find that the use of the methodology above these values is outside the scope of our review, and would require further justification and NRC review."Turkey Point does not have UPI and thus this SER requirement is N/A Turkey Point Units 3 and 4 EPU Licensing Report App. A-24 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4TableA.9-2Best-Estimate Large Break LOCA - Usage Conditions1."A recommended justification for any future time step changes (first listed item). We require that W perform this justification as recommended, and retain traceable documentation of this action in its in-house plant records."This requirement is satisfied since all time step changes have been justified and documented in Westinghouse records.2."Based on Reference214 [

A.9-1], Attachment7, the analysis to determine the uncertainty distributions for accumulator and SI temperatures uses plant operating data and/or plant Technical Specifications. Therefore, this analysis must be performed for each plant."This requirement is satisfied since the analyzed accumulator and SI temperature ranges use plant operating data.3."On CQD [A.9-2] page7-24, Westinghouse stated the fuel pellet thermal expansion model in MATPRO-11, Revision1, Reference176 [

A.9-3], was simplified by omitting the corrections for molten fuel and mixed oxide (Pu). In Reference214 [

A.9-1], ListII, Item6, Westinghouse committed to resubmitting the relevant

WCOBRA/TRAC models for NRC review if the code will be used to analyze US licensed plants with molten fuel or mixed oxide."This requirement is satisfied since the PTN Large Break LOCA analysis does not support the use of molten fuel or mixed oxide.4."Westinghouse, in Reference214 [

A.9-1], ListII, Item8, committed to not changing the value and range of the broken loop cold leg nozzle loss coefficient for plant specific applications. Also, the values developed apply only to LBLOCA and must be justified for other applications."This requirement is satisfied since the range of the broken loop cold leg nozzle loss coefficient developed for LBLOCA was not changed for the PTN Large Break LOCA analysis.5."Westinghouse, in Reference214 [

A.9-1], Attachment9, gave additional explanation on its use of the full Method of Characteristics model for each time step in the code implementation of choked flow. In the above reference, Westinghouse committed to include the information in the CQD [A.9-2]."Westinghouse satisfied this requirement by adding the necessary text to the critical flow model description in Section4-8-2 of WCAP-12945-P-A and the ASTRUM topical report (WCAP-16009-P-A).6."Westinghouse noted that the choked flow solution is implemented in the pressure solution of the code rather than in the back substitution step after solving the pressure equation. This results in a smoother pressure and flow response in the code. In Reference214 [

A.9-1], Attachment9, Westinghouse committed to include this information in the CQD [A.9-2]."Westinghouse satisfied this requirement by adding the necessary text to the critical flow model description in Section4-8-2 of WCAP-12495-P-A and the ASTRUM topical report (WCAP-16009-P-A).

Turkey Point Units 3 and 4 EPU Licensing Report App. A-25 Safety Evaluation Report ComplianceTurkey Point Units 3 and 4L-2010-113Docket Nos. 50-250 and 50-251Attachment 4ReferencesA.9-1N. J. Liparulo, Westinghouse, letter to USNRC Document Control Desk, "Docketing of Supplemental Information Related to WCAP-12945-P," NSA-SAI-96-156, April30,1996.A.9-2S. M. Bajorek, et. Al., WCAP-12945-P-A, Volume1, Revision2, and Volumes2 through5, Revision1, "Code Qualification Document for Best Estimate LOCA Analysis," 1998.A.9-3D. L. Hagrman, G. A. Reymann, and R. E. Manson, MATPRO-Version11 (Revision1), A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, NUREG/CR-0497, Rev.1, 1980.7."Westinghouse, in Reference214 [

A.9-1], ListII Item10, committed to use the multiplier given in Reference214 [A.9-1], Attachment4, to account for rod-to-rod radiation effects in the heat transfer multiplier data base."Westinghouse applies a correction factor to the reflood heat transfer multipliers to account for rod-to-rod radiation effects, as described on page 25-5-26 of WCAP-12945-P-A. The same correction factor is applied with ASTRUM.TableA.9-2 (Continued)Best-Estimate Large Break LOCA - Usage Conditions

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