ML24305A000
| ML24305A000 | |
| Person / Time | |
|---|---|
| Site: | 99902078 |
| Issue date: | 10/31/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL/NRLB |
| References | |
| Download: ML24305A000 (18) | |
Text
From:
Getachew Tesfaye Sent:
Thursday, October 31, 2024 12:09 AM To:
Request for Additional Information Cc:
Thomas Hayden; Mahmoud -MJ-Jardaneh; Griffith, Thomas; Sfowler@nuscalepower.com; Bode, Amanda; NuScale-SDA-720RAIsPEm Resource
Subject:
Nonproprietary - NuScale Non-LOCA TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology Revision 4 - Request for Additional Information No. 032 (RAI-10297-R1)
Attachments:
TR-0516-49416 (Non-LOCA) - RAI-10297-R1 - FINAL NON-PROPRIETARY.pdf Attached please find NRC staffs nonproprietary request for additional information (RAI) concerning the review of NuScale Topical Report TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology Revision 4 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23005A305). The encrypted proprietary version will be submitted in a separate email.
Please submit your technically correct and complete response by the agreed upon date to the NRC Document Control Desk.
If you have any questions, please do not hesitate to contact me.
Thank you, Getachew Tesfaye (He/Him)
Senior Project Manager NRC/NRR/DNRL/NRLB 301-415-8013
Hearing Identifier:
NuScale_SDA720_RAI_Public Email Number:
42 Mail Envelope Properties (BY5PR09MB56829F8CFE08FB8A0C6413288C552)
Subject:
Nonproprietary - NuScale Non-LOCA TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology Revision 4 - Request for Additional Information No. 032 (RAI-10297-R1)
Sent Date:
10/31/2024 12:08:40 AM Received Date:
10/31/2024 12:08:46 AM From:
Getachew Tesfaye Created By:
Getachew.Tesfaye@nrc.gov Recipients:
"Thomas Hayden" <Thomas.Hayden@nrc.gov>
Tracking Status: None "Mahmoud -MJ-Jardaneh" <Mahmoud.Jardaneh@nrc.gov>
Tracking Status: None "Griffith, Thomas" <tgriffith@nuscalepower.com>
Tracking Status: None "Sfowler@nuscalepower.com" <sfowler@nuscalepower.com>
Tracking Status: None "Bode, Amanda" <abode@nuscalepower.com>
Tracking Status: None "NuScale-SDA-720RAIsPEm Resource" <NuScale-SDA-720RAIsPEm.Resource@nrc.gov>
Tracking Status: None "Request for Additional Information" <RAI@nuscalepower.com>
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BY5PR09MB5682.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 693 10/31/2024 12:08:46 AM TR-0516-49416 (Non-LOCA) - RAI-10297-R1 - FINAL NON-PROPRIETARY.pdf 192600 Options Priority:
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1 REQUEST FOR ADDITIONAL INFORMATION No. 032 (RAI-10297-R1)
BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology Revision 4, ISSUE DATE: 10/31/2024
=
Background===
By letter dated October 31, 2023, NuScale Power, LLC (NuScale or the applicant) submitted Part 2, Final Safety Analysis Report (FSAR) of the NuScale Standard Design Approval Application (SDAA) for its US460 standard plant design (Agencywide Documents Access and Management System Accession No. ML23306A033). FSAR Chapter 15, Transient and Accident Analyses (ML23304A365) references the accident analysis methodology presented in NuScale topical report (LTR) "Non-Loss-of-Coolant Accident Analysis Methodology," TR-0516-49416-P, Revision 4 (ML23005A306), submitted January 5, 2023 (Non-LOCA LTR). The applicant submitted the US460 plant SDAA in accordance with the requirements of Title 10 Code of Federal Regulations (10 CFR) Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, Subpart E, Standard Design Approvals. The NRC staff has reviewed the information presented in TR-0516-49416-P, Revision 4, and determined that additional information is required to complete its review.
Regulatory Basis General Design Criterion (GDC) 1, Quality standards and records, and GDC 30, Quality of reactor coolant pressure boundary, require RCS components design, fabrication, erection, and testing to meet the highest quality standards practical.
GDC 5, Sharing of structures, systems and components, in 10 CFR Part 50, Appendix A, requires that any sharing among nuclear power units of structures, systems, and components (SSCs) important to safety will not significantly impair their safety function.
GDC 10, Reactor Design, requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLS) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).
GDC 13, Instrumentation and control, requires that instrumentation be provided to monitor variables and systems over anticipated ranges for normal operations, for AOOs, and for accident conditions and that controls be provided to maintain these variables and systems within prescribed operating ranges.
GDC 15, Reactor coolant system design, requires that the reactor coolant system (RCS) and its associated auxiliaries shall be designed with sufficient margin to assure that the design conditions of the pressure boundary are not exceeded during normal operations, including AOOs.
GDC 17, Electric power systems, requires that an onsite and offsite electric power system shall be provided to permit the functioning of SSCs important to safety. The safety function for each system (assuming the other system is not working) shall be to provide sufficient capacity and capability to ensure that the acceptable fuel design limits
2 and the design conditions of the reactor coolant pressure boundary (RCPB) are not exceeded as a result of an AOO and that core cooling, containment integrity, and other vital functions are maintained in the event of an accident.
GDC 20, Protection system functions, requires that the reactor protection system shall be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that the plant does not exceed SAFDLs during any condition of normal operation, including AOOs.
GDC 25, Protection system requirements for reactivity control malfunctions, requires that the reactor protection system shall be designed to ensure that SAFDLs are not exceeded for any single malfunction of the reactivity control system, such as accidental withdrawal of control rods.
GDC 26, Reactivity control system redundancy and capability, requires that a reactivity control system shall be provided that is capable of reliably controlling reactivity changes to assure that SAFDLs are not exceeded even during AOOs. This is accomplished by ensuring that the applicant has allowed an appropriate margin for malfunctions such as stuck rods.
GDC 27, Combined reactivity control systems capability, requires that reactivity control systems shall be designed with the combined capability of reliably controlling reactivity changes to assure that, under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained.
GDC 28, Reactivity limits, requires that reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither: (1) result in damage to the RCPB greater than limited local yielding nor (2) sufficiently disturb the core, its support structures, or other reactor pressure vessel (RPV) internals to impair significantly the capability to cool the core.
GDC 31, Fracture prevention of reactor coolant pressure boundary, requires that the RCS shall be designed with sufficient margin to ensure that the boundary behaves in a nonbrittle manner, and that the probability of propagating fracture is minimized.
GDC 34, Residual heat removal, requires that a system shall be provided with the capability to transfer decay heat and other residual heat from the reactor so that SAFDLs and pressure boundary design limits are not exceeded.
3 Question NonLOCA.LTR-1 Issue The Non-LOCA methodology does not provide an adequate basis for model validation vs. test data ((. Riser holes have an impact on thermal hydraulic conditions, for example but not limited to, initial conditions, natural circulation and RCS response; these impacts and any others need to be addressed. Information Requested a) Provide justification via sensitivity studies demonstrating that NRELAP5 can adequately calculate the actual expected NPM response with respect to the integrated topical report model. The total integral response of the Non-LOCA LTR model (from beginning of the event until the end of the period of interest) to various design basis events and conditions should be validated, (( }}. The potential impact on the RCS response and conditions for the integrated model on the figures of merit calculated by the topical report model and downstream activities should be shown analytically as described below:
- i.
Provide sensitivity analyses and evaluations of the impact of riser holes on the integral effects tests responses used for Evaluation Model validation for the Non-LOCA LTR, including the referenced integral effects test responses. The various impacts of riser holes on the integral test validation response should be captured in the impacts on the integrated NRELAP5 evaluation model results for Non-LOCA analyses. ii. Provide sensitivities and evaluations for the impact of riser holes on the total integral response of the LTR model (Non-LOCA) to various design basis events and conditions that show that the test validation response to riser holes is captured. The impact on the RCS response and conditions for the integrated model on the figures of merit calculated by the Non-LOCA LTR model should be shown analytically. b) Revise the LTR to include the above information.
4 Question NonLOCA.LTR-31, 32, 46, 56, 65 Issue The methodologies presented in the Non-LOCA LTR rely on sensitivity studies for various parameters in various events as identified in Table 7-7, Table 7-14, Table 7-19, Table 7-24, Table 7-28, Table 7-32, Table 7-36, Table 7-40, Table 7-44, Table 7-48, Table 7-52, Table 7-56, Table 7-60, Table 7-64, Table 7-68, Table 7-70, 7-74, Table 7-82, Table 7-86, and Table 7-91. However, the LTR only indicates what parameters require sensitivity studies without providing methodologies for these sensitivity studies. Additionally, the aforementioned tables also indicate bias directions for some parameters that result in limiting margins to figures of merit but does not explain how the biasing methodology considers the full range of operating conditions for each transient to determine the minimum margin parameters acceptance criterion (e.g. primary pressure). Determination of the limiting values of the input parameters The Non-LOCA LTR does not provide a complete methodology to determine the limiting values of these parameters to assure the analyses will appropriately capture limiting initial conditions that would be challenging to acceptance criteria. It is unclear how these sensitivity studies are to be performed as part of the methodology. While the audited documents (ML23067A300) provided sample calculations that showed how the sensitivity studies are performed for these audited cases, the methodology used in the calculation may not be appropriate for all parameters as indicated in some event-specific methodology descriptions in the Non-LOCA LTR (e.g. Non-LOCA LTR Section 7.2.14.1). Determination of the initial points for which sensitivity studies are to be performed Many events require consideration of a spectrum of event initiators. The purpose of sensitivity studies is to find the gradient of the function, which indicates the direction and magnitude the function would respond to a small perturbation. The results of the sensitivity analyses depend on the initial points at which the analyses are performed. It is important to identify those points to appropriately assess sensitivities to the given parameters. However, the Non-LOCA LTR does not describe how the points along this spectrum are selected for analysis, or how other parameters are varied with these points. (( }} but is not clear how these points will be identified and whether this process will be repeated when other parameters are varied. The previous revision of this LTR contained example sensitivity studies and accompanying discussion that demonstrated how sensitivity studies were to be performed, but these example sensitivity studies and discussion were removed in the current revision. As such, the Non-LOCA LTR does not adequately describe the methodology for ensuring that limiting cases are identified. When sensitivity studies should be performed During a regulatory audit (ML23067A300), NRC staff requested clarification of various statements in Section 7.2 of the Non-LOCA LTR that sensitivity studies are performed as needed to identify the limiting responses for acceptance criteria challenged by this event.
5 (( }}. However, the Non-LOCA LTR does not specify when sensitivity studies are needed, and instead leaves it to the discretion of the analysts. Generalized conclusion RCS pressure is insensitive to SG heat transfer During its review, the staff also notes that for several specific events (e.g., turbine trip and main steam isolation valve (MSIV) closure) the initial conditions, bias, and conservatisms tables, Table 7-32 and Table 7-40, denote the steam generator heat transfer is set as normal. The staff requested the applicant to provide information to support that these settings are appropriate. During the audit (ML23067A300), the applicant stated ((
}}.
Information on what additional sensitivity analyses are needed The methodology presented in the Non-LOCA LTR states that additional sensitivity studies are performed on parameters, as necessary, for multiple events, such as a decrease in feedwater temperature, increase in feedwater flow, etc., to identify the case(s) with a potentially limiting MCHFR. For example, for the turbine trip / loss of external load event, Section 7.2.6.1 of the Non-LOCA LTR states: Sensitivity studies on initial primary temperature and primary/secondary pressures are performed to identify the conditions that maximize peak primary and secondary pressures. Additional sensitivity studies are performed on other parameters, as necessary, to identify the case(s) with the potentially limiting peak primary and secondary pressures. The Non-LOCA LTR, however, does not provide information on how to do the additional sensitivity studies, what sensitivities will be examined in these additional studies, and what the acceptance criterion is for these sensitivity studies, i.e., when a sensitivity study is sufficient to identify the limiting cases or how the analyst would determine the limiting case has been correctly identified. Supporting information on various biasing schemes In all of the event-specific tables showing initial conditions, biases, and conservatisms, the staff notes that biasing the initial conditions for the normal initial full power case will not always result in a conservative change in the acceptance criterion parameters, due to the potential non-linear response from a change in a parameter (e.g., the coupled neutronic and thermal-hydraulic responses) and event progression, particularly when assuming different initial conditions in the base or reference case for an event. For example, the relative power peaking within an assembly may skew to significantly higher values at lower power levels than the one at full power, which can challenge the acceptance criteria in an AOO and can be more limiting than at full power. (( }}. In determining a conservative bias, those cases which do have a maximum relative change in acceptance criteria need to be considered if they become the limiting case in determining the minimum margin to acceptance criteria. The applied biases also need to account for the potential non-linear response of a parameter where the most
6 conservative case could be missed. Analyses to demonstrate that the biases used are appropriate and conservative are needed. Applicability of sensitivity studies performed for Non-LOCA LTR Revision 3 During the audit (ML23067A300), NuScale indicated that there are no substantial changes in Non-LOCA event progressions or the important phenomena between the NPM-160 and NPM-20 and that the use of biases in Revision 3, or update of biases based on NPM-160 insights, is reasonable. However, the LTR does not include justification for the applicability of the methodologies to the NPM-20 design, which includes significant changes in the design features of the NPM-20 reactor. Information Requested a) Modify the LTR to provide methodologies for performing sensitivity studies for the parameters that are identified as varied in Table 7-7, Table 7-14, Table 7-19, Table 7-24, Table 7-28, Table 7-32, Table 7-36, Table 7-40, Table 7-44, Table 7-48, Table 7-52, Table 7-56, Table 7-60, Table 7-64, Table 7-68, Table 7-70, 7-74, Table 7-82, Table 7-86, and Table 7-91 as well as the methodology for biasing parameters in these tables. The information should include:
- i.
How to perform the sensitivity studies required by the LTR methodology ii. How to determine the initial conditions at which the parameters are to be varied iii. The acceptance criteria iv. How biases are assured to be conservative at the other statepoints examined in sensitivity studies. b) Modify the LTR to provide methodologies for performing sensitivity studies for the parameters that are referred to as as needed or as necessary in Section 7 of the LTR. The information should include:
- i.
Under what conditions the sensitivity studies need to be performed ii. How to perform the sensitivity studies required by the LTR methodology iii. How to determine the initial conditions at which the parameters are to be varied iv. The acceptance criteria
- v.
How biases are assured to be conservative at the other statepoints examined in sensitivity studies. c) Modify the LTR to provide methodologies for performing the additional sensitivity studies as described in Section 7 of the LTR. The information should include:
- i.
Under what conditions additional sensitivity studies need to be performed ii. How to perform the sensitivity studies required by the LTR methodology iii. How to determine the initial conditions at which the parameters are to be varied iv. The acceptance criteria
7
- v.
Biases are assured to be conservative at the other statepoints examined in sensitivity studies. d) Revise the LTR to:
- i.
Clearly identify which evaluation methodology (EM), as approved in revision 3 of the LTR, remains applicable to which event(s) as listed in Section 7 in revision 4 of the LTR ii. Provide technical justifications supporting the EM applicability conclusion made for each transient event. The design changes from NPM-160 to NPM-20 must be addressed in the technical justifications.
8 Question NonLOCA.LTR-3, 9, 18, 19, 20, 21, 69 Issues The methodology in NuScale topical report "Loss-of-Coolant Accident Evaluation Model," TR-0516-49422-P, Revision 3 (ML23008A001) (LOCA LTR) and the Non-LOCA LTR does not contain sufficient scaling analysis updates with the NIST-2 testing facility to capture higher power operations of the NPM-20 RCS, containment, and cooling pool. Between the LOCA LTR and Non-LOCALTR there is missing justification to extend the similarity of the NPM-160/NIST-1 to the NPM-20/NIST-2 in terms of the distortion analysis of important phenomena. This scaling analysis is discussed in the justification for applicability of the NRELAP5 code but not provided in the submission adequately. Code validation for predicting DHRS performance The Non-LOCA LTR presents a validation of the NRELAP5 code using the NIST-2 test data. The code validation is extended to applications of the code for LOCA events (( }}. Similarities (or distortion) of NIST-2 tests for code validation Distortion analysis is missing in the Non-LOCA LTR for validation of the NRELAP5 code based on the NIST-2 test for analyses of Non-LOCA and LOCA events. A distortion analysis must be performed to identify the necessary correction factor in the code and models because comparison between the NIST-2 test facility and the NPM-20 design shows significant dissimilarities. Integral and separate effects tests are used to justify the development and applicability of an evaluation model to a given design. Since test facilities are not typically full-scale, distortions exist that can affect local and global elements of the analysis when compared to the full-scale plant. Therefore, a scaling analysis needs to be performed that identifies important non-dimensional parameters related to geometry and key phenomena and scaling distortions and their impact on the code assessment must be identified and evaluated. Applicability of the NIST-2 tests for validating code for analyzing events In the Non-LOCA LTR, the applicability of the NRELAP5 DHRS modeling capability is extended to the LOCA LTR (( }}.
9 ((
}}
DHRS condensate flow oscillations and nodalization (( }}. Prediction of the impact of coolant inventory on DHRS performance During audit discussions, NuScale presented information that showed performance degradation in certain DHRS loop inventory ranges. The SG-DHRS loop inventory is determined when the SG-DHRS loop is isolated for DHRS operation. In various Chapter 15 events, the loop isolation timing varies depending on the event progression and plant response. In reviewing the NRELAP assessment, NuScale identified the influence of DHRS loop inventory on heat removal. NuScale made available for staff audit (ML23067A300) several studies to address the staffs concern regarding the similarity between NIST-2 and the NPM-20. (( }}
10 ((
}}.
NRELAP5 modeling of secondary-side mass inventory (( }}. Information Requested a) Provide NRELAP5 DHRS model validation for LOCA events (( }}. b) (( }}. Provide the bounding uncertainty of the DHRS methodology in LOCA until long-term cooling is established with or without ECCS operation. c) (( }}. d) (( }}. e) Explain the impact of (( }} the scaling analysis in the Non-LOCA tests and the scalability to NPM-20. f) (( }}. If not, provide a most-limiting SG-DHRS loop inventory analysis for applicable Chapter 15 events to ensure sufficient DHRS heat removal capacity.
11 Question NonLOCA.LTR-3, 18, 27 Issue The NPM-20 reactor pool is modeled by NuScale (( }}. In addition, the staff notes that FSAR Chapter 5 states: A turbine trip at full power without bypass capability is the most severe AOO and is the bounding event used in the determination of RSV capacity and the RPV overpressure analyses. Sizing of the RCS and the PZR steam space avoids an RSV lift during normal operational transients that produce the highest RPV
12 pressure at full power conditions, with system and core parameters within normal operating range. In the event of a safety valve lift, the size of the PZR steam space is sufficient to preclude liquid discharge. The analytical model used for the analysis of the overpressure protection system and the basis for its validity is in the NuScale Topical Reports "Non-Loss-of-Coolant Accident Analysis Methodology" and "Loss-of-Coolant Accident Evaluation Model". (( }}. Information Requested a) The Non-LOCA DHRS base model and its variants are not only used in Non-LOCA event simulations but also in LOCA analysis for FSAR Section 15.6.5 and Chapter 5 normal shutdown analysis. Provide evaluations and bases information that address the concerns, as described above, (( }}, NuScale is requested to describe, from the perspective of condenser-to-pool heat transfer, the startup of a DHRS loop from cold (reactor pool temperature) conditions to the point of peak heat transfer rate, with focus on the various condenser-to-pool heat transfer modes and wall heat transfer correlations that will be involved with the startup of DHRS. b) Provide an evaluation of the DHRS performance under a scenario (( }}.
13 Question NonLOCA.LTR-50 Issue By letter dated January 6, 2023 (ML23011A012), NuScale submitted NRELAP5 files via DVD to support review of the Non-LOCA LTR and other LTRs that support the SDAA. NuScale informed NRC in February 2024, that a code error was found. Subsequently, NuScale informed NRC that a new code version, NRELAP5 v1.7, was being released and would be used for the SDAA and supporting topical reports. By letter dated August 6, 2024 (ML24228A242), NuScale submitted (( }}. Information Requested In order for the staff to complete its technical review of the analytical approach and base its findings on submitted information relevant to the version of the code utilized in developing the EM, NuScale is requested to: a) Re-submit any of the remaining files impacted by this code version change. All files affected by the revision to v1.7, (( }}, need to be resubmitted to provide a complete submittal package for all four of these LTRs. (( }}. Note that similar requests are being made for all four LTRs, and responses can reference each other. b) The markup to Non-LOCA LTR draft revision 5 that the staff audited states "testing associated with the release of the Reference 31 version of NRELAP5 confirmed no impact on the Non-LOCA EM." Make available for staff audit the documentation of the completed testing. c) Make available for audit an evaluation of the new version of the NRELAP5 code with all code qualification and design basis event analyses where the choked flow is involved to demonstrate impact on the current calculations. Provide LTR markups with updates to describe choked flow modeling changes and add or replace all evaluations in the LTR that are impacted.
14 Question NonLOCA.LTR-55 Issue As a result of the NRC staffs Quality Assurance Program (QAP) implementation inspection in February 2024, the NRC staff understands that the Non-LOCA LTR is controlled as a licensing product within NuScale, and the document itself is not considered to be the Non-LOCA Evaluation Model (EM). It is the NRC staffs understanding that information presented in the Non-LOCA LTR is derived from engineering documents. As such, the NRC staff needs to determine if there is information presented or conclusions stated in the Non-LOCA LTR, which is drawn from engineering documents that are not subject to design verification consistent with 10 CFR 50 Appendix B in accordance with Part II, Quality Assurance Program Description Details, Section 2.3.1, Design Verification of Topical Report MN-122626, Revision 1-A, NuScale Power, LLC Quality Assurance Program Description (ML24033A318) (hereafter referred to as QAPD LTR). This includes information subject to Part III, Nonsafety-Related Structures, Systems, and Components (SSC) Quality Control, Section 3.1.3, Design Control in the QAPD LTR, regardless of whether the verification process has been conducted in accordance with this section of the QAPD LTR, since the Section 3.1.3 verification is not required to conform to 10 CFR 50 Appendix B. Information Requested a) Provide a list of any such documents that have not undergone design verification as described above. For any of these documents not already available in eRR for staff audit, provide those additional documents available for audit. NuScale may choose to provide any relevant information regarding which portions of the LTR or conclusions those documents support. b) In addition, provide a listing of any section of the Non-LOCA LTR that are either not considered part of the EM or not directly supported by engineering documents that are part of the EM.
15 Question NonLOCA.LTR-60 Issue Section 7.1.2, Treatment of Plant Controls, of TR-0516-49416-P, Non-Loss-of-Coolant Accident Analysis Methodology, describes the approach for determining whether operation of non-safety-related control systems should be modeled. Specifically, the LTR states that operation of the control system will be modeled if it leads to a more severe event response and will not be modeled if operation leads to a less severe event response. Topical report section 7.2.2.3, Biases, Conservatisms, and Sensitivity Studies, identifies that as the increase in feedwater flow event creates the potential for overfilling the steam generator, additional sensitivity studies are performed to identify the case(s) with a potentially limiting steam generator level. (( }}. As described in Section 5.4.3.3.4 of the final safety analysis report (FSAR) provided with the NPM-20 standard design approval application, high DHRS inventory may inhibit DHRS performance if there is inadequate surface area in contact with the two-phase mixture for boiling and condensation to be effective. In response to questions raised during the audit, NuScale provided the biases, conservatisms, and sensitivity studies that will be evaluated to identify the highest steam generator level resulting from a potential increase in feedwater flow event. (( }}. NRC staff audited NuScales calculation of the increase in feedwater flow event supporting the standard design approval application for the NPM-20 design. (( }}. (( }}. While NRC staff notes that the calculation methodology contains some conservatisms which help ensure a bounding steam generator level is identified (( }} it is unclear to NRC staff if these conservatisms compensate for the
16 potential nonconservatism from assuming (( }}. Information Requested Demonstrate that each of these assumptions result in the limiting steam generator level or limiting DHRS capacity for the increase in feedwater flow event, or revise the Non-LOCA LTR with a revised treatment of (( }}. If assuming (( }} is more limiting for the NPM-20 design, revise the Non-LOCA LTR as well and make corresponding changes to FSAR Chapters 5, 15, or others as appropriate.}}