ML24346A181
Text
Response to SDAA Audit Question Question Number: A-4.4-1 Receipt Date: 03/23/2023 Question:
Section 4.4 of the SDAA is missing sufficient information related to the subchannel analysis necessary for the staff to perform its review. SDAA Section 4.4 is missing:
a.
a discussion of the subchannel analysis b.
the analytical results with maps of the typical distributions for the following thermal hydraulic parameters: radial power distribution, MCHFR, maximum rod clad outer wall temperature, maximum rod heat flux, average channel mass flux, maximum channel equilibrium quality, and exit channel fluid temperature c.
geometries (volumes, flow areas, and volume lengths) of reactor coolant system components used for calculating RCS loop flow.
d.
information related to NuScale modified correlations. This information should be similar to that provided for NSP4 and NSPN-1 correlations.
e.
application specific inputs, including uncertainties, to the Statistical CHFR analytical limit (SCHFAL), including the applicability range for the conditions that were used in determining the SCHFAL.
f.
application specific inputs for NSPN-1.
Response
Response to subpart a.
The final safety analysis report (FSAR) Section 4.4.4 describes the evaluation methodology and analysis tools utilized for the subchannel analysis. The subchannel steady-state results provided in Section 4.4 and the transient results in Chapter 15 are performed in accordance with the methodology defined in Reference 4.4-1 as supplemented by Reference 4.4-2 of Section 4.4.
Section 4.4.4.5.2 provides explicit discussion and pointers related to the application of the subchannel analysis model.
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Uncertainties that affect the prediction of CHF are incorporated into the CHF analysis limit determined using an NRC-approved methodology, the event-specific transient analysis, or a combination of the analysis limit and transient analysis. FSAR Section 4.4.2.7 and Table 4.4-2 indicate where uncertainties are addressed for the standard design approval application.
Additional details of the uncertainties and their application is provided in response to subpart e.
Section 4.4.3.3 and Figure 4.4-3 provides a thermal margin limit map for steady-state operation demonstrating that CHF is precluded. Section 4.4.4.6 and Table 4.4-5 provide the peak fuel temperature and peak linear heat rate demonstrating margin to fuel melt. With the above, it is demonstrated that fuel integrity is maintained. Additional details of the steady-state subchannel results are provided in response to subpart b.
Response to subpart b.
Maps of typical thermal hydraulic parameters for steady state operation of the equilibrium cycle are provided in Figures 1 through 7. These figures are developed based on a fully detailed core subchannel model using best estimate inputs. The minimum critical heat flux ratio for steady-state operation for the equilibrium cycle is 4.9 which is higher than the statistical analysis limit of 1.33.
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Figure 1: Radial Power Distribution NuScale Nonproprietary NuScale Nonproprietary
Figure 2: Minimum Critical Heat Flux Ratio NuScale Nonproprietary NuScale Nonproprietary
Figure 3: Maximum Rod Clad Outer Wall Temperature (degrees F)
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Figure 4: Maximum Rod Heat Flux (MBtu/hr-ft2)
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Figure 5: Average Channel Mass Flux (Mlbm/hr-ft2)
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Figure 6: Maximum Channel Equilibrium Quality NuScale Nonproprietary NuScale Nonproprietary
Figure 7: Exit Channel Fluid Temperature (degrees F)
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Response to subpart c.
The geometries used for calculation of thermal hydraulic boundary conditions are shown in Table 1.
Table 1: Nominal Geometries of RCS Coolant System Components RCS Region Total RCS Region Volume (ft3)
RCS Sub-Region Description Average Flow Area (ft2)
Length (ft)
Riser 635 Lower riser and transition 24.9 9.4 Upper riser and turn 15.4 26.0 Downcomer 1199 Downcomer (including steam generators) 26.4 45.4 Core 89 Fuel assemblies 10.3 7.9 Reflector cooling channel 0.9 7.9 Pressurizer 577 Pressurizer heaters /
main steam plenums 36.1 1.7 Cylindrical pressurizer 61.3 6.9 Reactor pressure vessel top head 41.2 2.2 Response to subpart d.
FSAR Section 4.4.2.1 identifies the CHF correlations utilized. This section identifies three sets of models: (1) NSP4, (2) NSPN-1, and (3) industry-developed correlations. The NSP4 CHF correlation is used to evaluate thermal margin for normal operation, AOOs, IEs, and accidents.
Events exhibiting a rapid depressurization are evaluated using NSPN-1 and industry-developed CHF correlations as described in detail in FSAR Reference 4.4-3. The term industry-developed (NuScale modified) refers to modified Griffith-Zuber and Extended Hench-Levy correlations which are unchanged from the previous revision of the topical report (FSAR Reference 4.4-3) that was approved in support of NRC review of the design certification application (ML20189A644).
Section 6.11 of the loss-of-coolant accident (LOCA) topical report, TR-0516-49422-P, Revision 3 (FSAR Reference 4.4-3), describes the NSPN-1, modified Griffith-Zuber, and Extended Hench-Levy CHF correlations. The NSPN-1 CHF correlation, discussed in Section 6.11.5 of the LOCA topical report, is used for assessing the Phase 0 LOCA response.
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The ranges of applicability for the NSPN-1 correlation are provided in Table 6-8 of the LOCA topical report. The modified Griffith-Zuber and Extended Hench-Levy CHF correlations, discussed in Section 6.11.3 and 6.11.4 of the LOCA topical report, are used to analyze the other phases of rapid depressurization events.
Response to subpart e.
Application-specific inputs used to develop the statistical critical heat flux analysis limit (SCHFAL) with the NSP4 correlation are shown in Table 2. The corresponding range of applicability for the statistical analysis limit is provided in Table 3.
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(( }}2(a),(c) Response to subpart f. Final safety analysis report (FSAR) Section 4.4.2.5 indicates that thermal margin calculations for the non-loss-of-coolant accident analyses in Chapter 15 are performed using VIPRE-01 with the NSP4 critical heat flux correlation using input from NRELAP5. FSAR Sections 15.6.5 and 15.6.6 for events characterized by rapid depressurization utilize NRELAP5 for both transient and thermal margin calculations with the Extended Hench-Levy, modified Griffith-Zuber, and NSPN-1 CHF correlations. Event-specific inputs for the inadvertent operation of the ECCS are discussed in FSAR Section 15.6.6.3.2 and Table 15.6-15. Event-specific inputs for the LOCA analysis are discussed in FSAR Section 15.6.5.3.2 and Table 15.6-9. ((
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Revised Response Revision 0 Response to subpart b. Upon request, this audit response can be submitted on the docket. Response to subpart c. The nominal geometries of RCS components is added as FSAR Table 4.4-7 as shown in the attached FSAR markup. Response to subpart d. Non-proprietary Extended Hench-Levy critical heat flux correlation information is added to Section 4.4 as shown in the attached FSAR markup. NuScale Nonproprietary NuScale Nonproprietary
Response to subpart e. The models and values that are used as input into calculating the statistical critical heat flux analysis limit as described in limitation and condition #2 in the advance safety evaluation report for TR-108601-P are defined in EC-105077, Revision 2, CHF Analysis Limit for Statistical Subchannel Analysis Applications. EC-105077 Rev. 2 is provided in the eRR as part of this response. Below is a guide to the specific sections of EC-105077, Rev. 2 that provide the specific values and justification or basis for the submodels listed in limitation and condition #2 of TR-108601-P: a. The maximum hot rod radial peaking analysis limit, including measurement uncertainties. (( }}2(a),(c) b. The models and values used to determine the core thermal power. (( }}2(a),(c) c. The models and values used to determine the core inlet flow. (( }}2(a),(c) d. The models and values used to determine the core inlet temperature. (( }}2(a),(c) e. The models and values used to determine the core exit pressure. (( }}2(a),(c) f. The models and values used to determine the enthalpy rise measurement uncertainty. EC-105077 Rev. 2 - Section 3.3 and 4.2.2 g. The models and values used to determine the enthalpy rise engineering uncertainties. EC-105077 Rev. 2 - Section 3.3 and 4.2.2 NuScale Nonproprietary NuScale Nonproprietary
h. The models and values used to determine the heat flux engineering uncertainty. This is given in equation 3-16 of the SSAM. EC-105077 Rev. 2 - Section 4.2.3 i. The models and values used to determine the fuel rod bow uncertainty. This is given in equation 3-17 of the SSAM. EC-105077 Rev. 2 - Section 4.2.4 The SDAA SCHFAL value is 1.45 as reflected in the analysis result tables in Chapter 15. The SCHFAL is added to Section 4.4 as shown in the attached FSAR markup. The values reported in Table 2 and Table 3 of the original audit response were updated to reflect staff feedback during the review of TR-108601-P. Table 4 below provides the most recent application-specific inputs used to develop the statistical critical heat flux analysis limit.(( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
(( }}2(a),(c) (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
Revised Response Revision 1 TR-169856, Revision 0, NuScale US460 Statistical Subchannel Critical Heat Flux Analysis Probabilistic Uncertainties, nonproprietary file (ML24213A316) and proprietary file (ML24213A317) was submitted on July 31st, 2024 via LO-172648 - NuScale Power, LLC Submittal of NuScale Technical Report, NuScale US460 Statistical Subchannel Critical Heat Flux Analysis Probabilistic Uncertainties, TR-169856, Revision 0. Attached are FSAR markups incorporating TR-169856, Revision 0, by reference in FSAR Section 1.6 and Section 4.4. Markups of the affected changes, as described in the response, are provided below: NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-3 Draft Revision 2 of conditions. Additional details on the CHF correlation development is provided in Section 4.4.2.5. Audit Question A-4.4-1 The CHFR design limit for the NSP4 correlation that corresponds to a 95/95 assurance level is 1.21 with domain-specific modifiers as specified in the conditions and limitations of the NRC safety evaluation in Reference 4.4-4. The NSP4 critical heat flux correlation is used to evaluate thermal margin for normal operation, AOOs, IEs, and accidents, with the exception of those characterized by rapid depressurization. Events exhibiting a rapid depressurization are evaluated using the NSPN-1 and industry-developed CHF correlationsthe Extended Hench-Levy correlation described in Reference 4.4-3. The CHFR design limit for the NSPN-1 correlation that corresponds to a 95/95 assurance level is discussed in Reference 4.4-3. Audit Question A-4.4-1 The range of applicability of the NSP4 critical heat flux correlation is contained in Reference 4.4-4, and the ranges of applicability of the NSPN-1 and the Extended Hench-Levyindustry-developed critical heat flux correlations are contained in Reference 4.4-3. The transient response of the reactor system is dependent on the initial power distribution. Limits provided by the core design and the module protection system (MPS) ensure the NPM meets CHF design bases for AOOs. The technical specifications define the cycle-specific enthalpy rise hot channel factor and axial offset (AO) limits that must be maintained during operation. Section 4.3 provides additional discussion about the development and use of these limits. The MPS automatically initiates the protective actions necessary to mitigate the effects of the design-basis events identified in Table 7.1-1. The MPS reactor trip functions are listed in Table 7.1-3, including the associated process variables and analytical limits. The core design and thermal limits are developed such that the thermal margin criteria are not exceeded for normal operation and AOOs. Specifically, there is a 95/95 assurance that the hot rod in the core does not experience a CHF condition. For the purpose of this analysis, CHF is assumed to occur if the subchannel analysis-calculated CHFR is less than the allowable limit. For IEs and accidents, the total number of fuel rods that exceed the criteria are assumed to fail and are used in determining the radiological dose source term. 4.4.2.2 Linear Heat Generation Rate A limit on peak linear heat generation rate (PLHGR) is specified (Section 4.2) to ensure the fuel overheating SAFDL is not exceeded. The design limit on PLHGR maintains the fuel temperature below the centerline melt criterion.
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-5 Draft Revision 2 4.4.2.4 Core Pressure Drops and Hydraulic Loads Flow testing on a full-scale prototype fuel assembly was performed to establish flow component loss coefficients and other related flow characterization parameters for the fuel assembly. The form loss coefficients are used in the fuel assembly liftoff analysis and the subchannel analysis. Fuel assembly pressure drop tests were performed for a range of steady-state conditions as part of the CHF testing program. Pressure drops for the fuel assembly features are included in the subchannel model corresponding to their location in the fuel assembly as described in Reference 4.4-2. Section 4.2 confirms the fuel assemblies do not experience liftoff from the lower core plate under normal operating conditions and AOOs. 4.4.2.5 Correlations and Physical Data Audit Question A-4.4-1 Testing was performed in support of CHF correlation development. Development of the NSP4 and NSPN-1 critical heat flux correlations used data collected from testing at the Framatome Karlstein thermal hydraulic test loop (KATHY) facility on a fuel design prototypical of the NuFuel-HTP2TM fuel design (Reference 4.4-5 and Reference 4.4-3). The approved domain for the NSP4 critical heat flux correlation was extended using data obtained from testing at Stern Laboratories on a fuel design similar to the fuel design, with the exception that it did not include flow mixing grids (Reference 4.4-4). The Extended Hench-Levy correlation (Reference 4.4-3) is based on Stern data with the KATHY data providing validation to conservatively predict NuFuel HTP2' critical heat flux performance for an inadvertent opening of a reactor pressure vessel valve event. Additional discussion of the CHF test programs is provided in Section 1.5.1. Comparisons between Stern Laboratories and KATHY data can be found in Reference 4.4-5 and Reference 4.4-4. The limiting non-loss-of-coolant accident analyses in Chapter 15 are performed using the NRELAP5 code. Once the limiting cases for each transient are identified, the thermal margin is determined using the VIPRE-01 subchannel core model with the NSP4 critical heat flux correlation analysis limit. Audit Question A-4.4-1 As discussed in Section 15.6.1 and Section 15.6.6, respectively, the analyses of an inadvertent opening of a reactor safety valve and of an inadvertent operation of the emergency core cooling system are performed using the NRELAP5 code. The loss-of-coolant accident analysis discussed in Section 15.6.5 is also performed using NRELAP5. Thermal margin for these events is determined using the NRELAP5 code with the NSPN-1 and the Extended Hench-Levy correlationindustry-developed CHF correlations as detailed in Reference 4.4-3.
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-8 Draft Revision 2 neutronic code uncertainty accounted for in the core design, no radial power distribution penalty needs to be applied in the subchannel analysis. Fuel rod bowing can have a negative impact on CHF because of reduced flow area in the hot channel. The penalty is applied conservatively based on the methodology for rod bow in Reference 4.4-8, which is demonstrated to be applicable for the NuScale fuel design in Reference 4.4-9 and Reference 4.4-10. The core inlet flow distribution is discussed in Section 4.4.2.3. For the subchannel analysis methodology, inlet flow distribution uncertainty is applied to the hot or limiting assembly. The open lattice of the core allows flow redistribution to occur for inlet flow imbalances and the applied flow reduction to the hot assembly has a minimal effect on MCHFR. The open upper plenum design allows for pressure equilibrium and no core exit pressure distribution uncertainty is necessary for the subchannel analyses. 4.4.2.7.4 Critical Heat Flux Analysis Limit Audit Question A-4.4-1 Utilizing the NSP4 CHF correlation, the uncertainties identified in Table 4.4-2 are combined using the statistical method described in Reference 4.4-2. As reflected in the analyses of Chapter 15, the resulting analysis limit of 1.45 includes domain-specific modifiers in accordance with the conditions and limitations in the NRC safety evaluation for Reference 4.4-4. The corresponding range of applicability for the statistical analysis limit is provided in Table 4.4-8. Reference 4.4-3 describes the application of uncertainties to the NSPN-1 critical heat flux correlation resulting in an analysis limit of 1.20. 4.4.2.8 Flux Tilt Considerations Radial tilt is a condition where the power is not symmetric between azimuthally symmetric fuel assemblies. Azimuthal power tilt is an allowable limit on operation. Once the flux tilt is beyond an allowable threshold, actions are required to remedy the condition. The design FH safety limit inherently accounts for the radial tilt, expressed as: Eq. 4.4-2
- where,
= technical specifications enthalpy rise design peaking factor, F
TS H FH 1 Tq + ( ) = F TS H
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-12 Draft Revision 2 limit during normal operation and AOOs, ensuring the SAFDLs are satisfied and fuel cladding integrity is demonstrated. For IEs and accidents, the total number of fuel rods that exceed the criteria and are assumed to fail is used as input for radiological dose calculation purposes. 4.4.4.1 Critical Heat Flux Correlation The functional forms of the NSP4 and NSPN-1 critical heat flux correlations are expressed as empirical relations dependent on a number of physical parameters including: pressure cold wall factor boiling length local mass flux local equilibrium quality The coefficients of the NSP4 critical heat flux correlation are determined with a cross-validation process and linear least-squares regression based on local condition parameters calculated with the VIPRE-01 subchannel thermal-hydraulics code. The form of the equation and correlation coefficients and the details of the development of the correlation are provided in Reference 4.4-5. The coefficients of the NSPN-1 critical heat flux correlation are determined using the same process as used for NSP4 but with local conditions calculated using NRELAP5. The form of the equation and correlation coefficients with details of the development for the NSPN-1 critical heat flux correlation are provided in Reference 4.4-3. 4.4.4.2 Core Hydraulics Audit Question A-4.4-1 As discussed in Section 4.4.2.3.1, a uniform inlet temperature distribution is applied. Table 4.4-3 lists the principal flow elements in the RPV flow path and describes the flow path. The geometries used for calculation of thermal hydraulic boundary conditions are summarized in Table 4.4-7. 4.4.4.3 Influence of Power Distribution The subchannel analysis basemodel is developed to bound a cycle-specific core as described in Reference 4.4-2. The model preserves limiting core conditions along with the operational envelope specified in the technical specifications. The envelope accounts for the power distribution throughout the core using design peaking factors and axial power shapes in combination with the limiting RCS parameters such as flow and pressure.
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-31 Draft Revision 2 Audit Question A-4.4-1 Table 4.4-7: Geometries of Reactor Coolant System Components RCS Region Total RCS Region Volume (ft3) RCS Sub-Region Description Average Flow Area (ft2) Length (ft) Riser 635 Lower riser and transition 24.9 9.4 Upper riser and turn 15.4 26.0 Downcomer 1199 Downcomer (including steam generators) 26.4 45.4 Core 89 Fuel assemblies 10.3 7.9 Reflector cooling channel 0.9 7.9 Pressurizer 577 Pressurizer heaters / main steam plenums 36.1 1.7 Cylindrical pressurizer 61.3 6.9 Reactor pressure vessel top head 41.2 2.2
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-32 Draft Revision 2 Audit Question A-4.4-1 Table 4.4-8: Statistical Subchannel Analysis Limit Range of Applicability Parameter Applicability Range Pressure (psia) 1800 to 2300 Inlet Temperature (degrees F) 431 to 568 Core inlet mass flux (Mlbm/hr-ft2) 0.21 to 0.7 Core local mass flux (Mlbm/hr-ft2) 0.20 to 0.721 Local equilibrium quality < 68%
NuScale Final Safety Analysis Report Material Referenced NuScale US460 SDAA 1.6-4 Draft Revision 2 Audit Question A-4.4-1, Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 Table 1.6-2: NuScale Referenced Technical Reports Incorporated by Reference Report Number Title Specific Report Sections Incorporated by Reference FSAR Section TR-118318, Revision 1 NuScale Design of Physical Security Systems All 9.5, 13.6, 14.2 TR-122844-P, Revision 0 NuScale Instrument Setpoint Methodology Technical Report All 7.0, 7.2 TR-121353-P, Revision 0 NuScale Comprehensive Vibration Assessment Program Analysis Technical Report All 3.9, 5.4 14.2 TR-121515-P, Revision 0 NuScale Power Module Seismic Analysis All 3.7, 3.12 TR-130877-P, Revision 0 Pressure and Temperature Limits Methodology All 5.2, 5.3 TR-121517-P, Revision 0 NuScale Power Module Short-Term Transient Analysis All 3.9 TR-130721-P, Revision 0 Use of Austenitic Stainless Steel for US460 Standard Design Reactor Pressure Vessel All 5.2, 5.3 TR-121354-P, Revision 0 NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report All 3.9, 14.2 TR-121507-P, Revision 0 Pipe Rupture Hazards Analysis All 3.6 TR-118976-P, Revision 0 Fluence Calculation Methodology and Results Section 6, Summary and Conclusions 4.3, 5.3 TR-117605-P, Revision 0 NuFuel-HTP-2 Fuel and Control Rod Assembly Designs Section 4, Design Evaluation, and Section 6, Control Rod Assembly 4.2, 4.3 TR-169856-P, Revision 0 NuScale US460 Statistical Subchannel Critical Heat Flux Analysis Probabilistic Uncertainties All 4.4
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-8 Draft Revision 2 neutronic code uncertainty accounted for in the core design, no radial power distribution penalty needs to be applied in the subchannel analysis. Fuel rod bowing can have a negative impact on CHF because of reduced flow area in the hot channel. The penalty is applied conservatively based on the methodology for rod bow in Reference 4.4-8, which is demonstrated to be applicable for the NuScale fuel design in Reference 4.4-9 and Reference 4.4-10. The core inlet flow distribution is discussed in Section 4.4.2.3. For the subchannel analysis methodology, inlet flow distribution uncertainty is applied to the hot or limiting assembly. The open lattice of the core allows flow redistribution to occur for inlet flow imbalances and the applied flow reduction to the hot assembly has a minimal effect on MCHFR. The open upper plenum design allows for pressure equilibrium and no core exit pressure distribution uncertainty is necessary for the subchannel analyses. 4.4.2.7.4 Critical Heat Flux Analysis Limit Audit Question A-4.4-1 Utilizing the NSP4 CHF correlation, the uncertainties identified in Table 4.4-2 are combined using the statistical method described in Reference 4.4-2. Reference 4.4-12 provides the models and values of the statistical uncertainties applied in the analysis limit calculation. As reflected in the analyses of Chapter 15, the resulting analysis limit includes domain-specific modifiers in accordance with the conditions and limitations in the NRC safety evaluation for Reference 4.4-4. Reference 4.4-3 describes the application of uncertainties to the NSPN-1 critical heat flux correlation resulting in an analysis limit of 1.20. 4.4.2.8 Flux Tilt Considerations Radial tilt is a condition where the power is not symmetric between azimuthally symmetric fuel assemblies. Azimuthal power tilt is an allowable limit on operation. Once the flux tilt is beyond an allowable threshold, actions are required to remedy the condition. The design FH safety limit inherently accounts for the radial tilt, expressed as: Eq. 4.4-2
- where,
= technical specifications enthalpy rise design peaking factor, F
TS H FH 1 Tq + ( ) = F TS H
NuScale Final Safety Analysis Report Thermal and Hydraulic Design NuScale US460 SDAA 4.4-20 Draft Revision 2 4.4-4 NuScale Power, LLC, Applicability Range Extension of NSP4 CHF Correlation, Supplement 1 to TR-0116-21012-P-A, Revision 1, TR-107522-P-A, Revision 1. 4.4-5 NuScale Power, LLC, NuScale Power Critical Heat Flux Correlations, TR-0116-21012-P-A, Revision 1. 4.4-6 AREVA NP Inc., COBRA-FLX: A Core Thermal-Hydraulic Analysis Code Topical Report, ANP-10311P-A, Revision 0. 4.4-7 Bahadur, Sher, U.S. Nuclear Regulatory Commission, letter to Pedro Salas, AREVA NP, Inc., January 29, 2013, Agencywide Document Access and Management System (ADAMS) Accession No. ML13135A053. 4.4-8 AREVA Inc., Computational Procedure for Evaluating Fuel Rod Bowing, XN-75-32(P)(A), Supplement 1-4. 4.4-9 NuScale Power, LLC, Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P-A, Revision 1. 4.4-10 NuScale Power, LLC, Framatome Fuel and Structural Response Methodologies Applicability to NuScale, TR-108553-P-A, Revision 0. 4.4-11 NuScale Power, LLC, Evaluation Methodology for Stability Analysis of the NuScale Power Module, TR-0516-49417-P, Revision 1. Audit Question A-4.4-1 4.4-12 NuScale Power, LLC, NuScale US460 Statistical Subchannel Critical Heat Flux Analysis Probabilistic Uncertainties, TR-169856-P, Revision 0.}}