ML25009A002

LLC - Response to NRC Request for Additional Information No. 032 (RAI-10297 R1) on the NuScale Standard Design Approval Application

ML25009A002
Person / Time
Site:	05200050, 99902078
Issue date:	01/09/2025
From:	Shaver M NuScale
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML25009A001	List: ML25009A002
References
RAIO-177581
Download: ML25009A002 (1)

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RAIO-177581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com January 09, 2025 Docket No.52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Response to NRC Request for Additional Information No. 032 (RAI-10297 R1) on the NuScale Standard Design Approval Application

REFERENCE:

NRC Letter to NuScale, Request for Additional Information No. 032 (RAI-10297 R1), dated October 31, 2024 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).

The enclosure to this letter contains the NuScale response to the following RAI question from NRC RAI-10297 R1:

NonLOCA.LTR-31, 32, 46, 56, 65 is the proprietary version of the NuScale Response to NRC RAI No. 032 (RAI-10297 R1, Question NonLOCA.LTR-31, 32, 46, 56, 65). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 is the nonproprietary version of the NuScale response.

This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions, please contact Amanda Bode at 541-452-7971 or at abode@nuscalepower.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on Janaury 09, 2025.

Sincerely, Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC

RAIO-177581 Page 2 of 2 01/09/2025 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Distribution:

Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Thomas Hayden, Project Manager, NRC

NuScale Response to NRC Request for Additional Information RAI-10297 R1, Question NonLOCA.LTR-31, 32, 46, 56, 65, Proprietary Version : NuScale Response to NRC Request for Additional Information RAI-10297 R1, Question NonLOCA.LTR-31, 32, 46, 56, 65, Nonproprietary Version : Affidavit of Mark W. Shaver, AF-177583

RAIO-177581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Response to NRC Request for Additional Information RAI-10297 R1, Question NonLOCA.LTR-31, 32, 46, 56, 65, Proprietary Version

RAIO-177581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Response to NRC Request for Additional Information RAI-10297 R1, Question NonLOCA.LTR-31, 32, 46, 56, 65, Nonproprietary Version

Response to Request for Additional Information Docket: 052000050 RAI No.: 10297 Date of RAI Issue: 10/31/2024 NRC Question No.: 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 NuScale Nonproprietary NuScale Nonproprietary

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. (( 2(a),(c) 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.

(( }} 2(a),(c) 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 ((

}} 2(a),(c)

• Information on what additional sensitivity analyses are needed NuScale Nonproprietary NuScale Nonproprietary

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. ((

}} 2(a),(c) 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 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 NuScale Nonproprietary NuScale Nonproprietary

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 criteriav.

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, NuScale Nonproprietary NuScale Nonproprietary

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. NuScale Response: Parts a, b, and c To address these three parts of the request additional information (RAI), NuScale provides a revised markup to TR-0516-49416-P, Revision 4, Non-Loss-of-Coolant Accident Analysis Methodology. The markup also addresses the NRC concerns from the original audit questions (i.e., A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, and A-NonLOCA.LTR-65) associated with this RAI. The markup is attached to this response. Because the markup includes changes on top of those previously provided in the original response to some of these audit questions, a summary of the changes is provided for convenience as follows:

Deleted use of "as necessary" or as needed discussion of sensitivity studies. The phrases are either just deleted or else replaced with a reference to the table that identifies the sensitivity studies to perform.

Added minimum margin criteria for reactor coolant system (RCS) pressure, steam generator (SG) pressure, and minimum critical heat flux ratio (MCHFR) figures of merit. A result that goes below the minimum margin then requires additional sensitivity studies be performed to ensure the limiting case is identified.

Restored the example sensitivity study results (both tables and associated text discussion) from the previously approved topical report version (TR-0516-49416-P-A, Revision 3).

In the first section with representative sensitivity results (i.e., Section 7.2.1.3), added discussion for how the representative sensitivity studies define elements of the method. The general explanation in Section 7.2.13 is then applied to the remainder of the events. Examples include: o how to vary parameters, o how to determine if results of sensitivity studies are sensitive or insensitive, and o how to consider a spectrum of parameters (usually tied to event initiator) rather than a simple variation. NuScale Nonproprietary NuScale Nonproprietary

Note that the markup originally provided is revised based on NRC feedback to provide the following clarifications.

The role of the reactor safety valve lifts in determining limiting reactor coolant system pressures is clarified.

The discussion of sensitivity studies that are repeated is clarified.

The criteria and examples for performing sensitivities studies to more than one parameter at a time is clarified.

The role of engineering judgment in reviewing sensitivity study results is clarified. In addition, NuScale identified some further changes to TR-0516-49416-P, Revision 4, to ensure conservatism of the event-specific methodology in Section 7.2. The changes are listed below. These changes are shown in markups that are appended to the end of the other markups (i.e., these new changes are shown in the last seven pages of the attached markups).

The initial pressurizer level for decrease in feedwater temperature (Table 7-7), increase in feedwater flow (Table 7-14), and increase in steam flow (Table 7-19) events is modified from biased high to varied. The basis (( }}2(a),(c) The resultant methodology now includes more sensitivity cases than previously and, therefore, this change is conservative.

The steam pressure control for the increase in steam flow (Table 7-19) event is clarified (( }}2(a),(c) may be used to model the increase in steam flow initiating event. This is not a change in methodology, but simply a clarification of how the increase in steam flow can be modeled.

The turbine throttle valve aspect of steam pressure control for the steam line break (Table 7-24) event is modified from enabled to varied. The basis (( }}2(a),(c) Therefore, sensitivity studies to determine the effect of the turbine throttle valve treatment should be performed to determine the treatment that NuScale Nonproprietary NuScale Nonproprietary

results in the more limiting MCHFR. The resultant methodology now includes more sensitivity cases than previously and therefore, this change is conservative. Part d The RAI states that the LTR does not include justification for the applicability of the methodologies to the NuScale Power Module (NPM)-20 design, which includes significant changes in the design features of the NPM-20 reactor. The RAI does not identify any specific design changes of concern to the non-loss-of-coolant accident (non-LOCA) event progressions. NuScale disagrees with the implied assertion that the design features of the NPM-20 reactor are significant with respect to the high-ranked phenomena for non-LOCA events that affect the non-LOCA event progressions. Part d of the request in this focuses specifically on the event-specific methodologies described in Section 7.2 and therefore these are discussed specifically in this response. However, technical justification for the event-specific methodologies described in Section 7.2 inherently rely on the overall EM development and applicability assessment provided in Sections 3, 4, 5, and 6 of TR-0516-49416-P. NuScale followed the evaluation model development and assessment process (EMDAP) described in Regulatory Guide 1.203, Transient and Accident Analysis Methods, to develop the previously approved non-LOCA evaluation model (EM), as described in TR-0516-49416-P-A, Revision 3. NuScale used the EMDAP to extend the EM for the NPM-20 design. There are four elements of the EMDAP: 1. Establish requirements for EM capabilities The analysis purpose, transient class, figures of merit, systems, components, phases, geometries, fields and processes to be modeled are unchanged. A key part of this element is to identify and rank phenomena and processes. NuScale evaluated the design changes from the NPM-160 to NPM-20. The NPM changes in initial conditions affect the ranges of some system transient process parameters. These changes were evaluated as part of demonstrating EM adequacy (element 4). NuScale determined that the design changes did not result in any new or different high-ranked phenomena for the scope of the non-LOCA event progression from event initiation through establishment of stable decay heat removal system (DHRS) cooling. Therefore, in development of TR-0516-49416-P, Revision 4, NuScale made minor changes to update Section 3, Plant Design Overview. Discussion of RCS pressure biasing to minimize critical heat flux margin was modified in Section 4, Transient and Accident Analysis Overview, to be applicable to both the NPM-160 and NPM-20 designs. Other parts of Section 4, and the NuScale Nonproprietary NuScale Nonproprietary

Section 5.1 identification of high-ranked phenomena remain applicable to the NPM-160 and NPM-20 designs. 2. Assessment base development The assessment basis of the previously approved TR-0516-49416-P-A, Revision 3, remains relevant to demonstrate applicability of the NRELAP5 code to predict the non-LOCA high-ranked phenomena and event progressions; therefore these sections of the topical report were maintained. NuScale chose to perform additional non-LOCA DHRS integral effects tests (IETs) at the NIST-2 facility to extend the NRELAP5 validation basis. These tests were scaled to representative NPM-20 conditions as described in TR-0516-49416-P, Revision 4, Section 5.3.7.2.2. 3. Develop evaluation model The structure of the previously approved EM is unchanged and remains applicable for both the NPM-160 and NPM-20 designs (i.e., use of NRELAP5 to predict the system thermal-hydraulic response). Therefore, in development of TR-0516-49416-P, Revision 4, NuScale made minor changes to update the Section 6 NRELAP5 plant model description to clarify critical aspects of the model while emphasizing that the base model for a specific NPM design reflects the parameters and features of that design. The previously approved NRELAP5 closure models remain applicable for the NPM-20 design. NuScale chose to develop (( }}2(a),(c) for steam generator primary side heat transfer and provided justification of the model applicability as described in Section 5.3.6. This closure model provides consistent results to the previously approved model (i.e., (( }}2(a),(c) ) and therefore is applicable to both the NPM-160 and NPM-20 designs. 4. Assess evaluation model applicability NuScale performed bottom-up and top-down assessments of the EM applicability to the NPM-20 design. As summarized in the TR-0516-49416-P, Revision 4, Section 5.4, many of the high-ranked phenomena for non-LOCA transients are also addressed as part of the LOCA evaluation model. Based on gap analyses, the non-LOCA EM applicability focuses on the steam generator and DHRS heat transfer phenomena. Bottom-up assessment concluded that the dominant constitutive models for boiling/condensation wall heat transfer phenomena in the steam generator and DHRS were developed over wide ranges of conditions and apply to the NPM-20 condition ranges. Top-down scaling analyses demonstrated the consistency in the dominant PI groups for steam generator and DHRS operation between the NPM-2 and NIST-2, thereby supporting the conclusion that the use of NRELAP5 in the EM is applicable for NPM-20 non-LOCA integral analysis. NuScale Nonproprietary NuScale Nonproprietary

Therefore, contrary to the statement in the RAI, NuScale followed the EMDAP to justify applicability of the EM to the NPM-20 design and the technical bases for the justification are provided throughout TR-0516-49416-P, Revision 4. When preparing the event-specific methodology in Section 7.2 of TR-0516-49416-P, Revision 4, the starting point was TR-0516-49416-P-A, Revision 3. In general, the content in Section 7.2 of Revision 4 is similar to Revision 3 and therefore the technical basis is similar to that previously reviewed and approved by the NRC. However, for each event in Section 7.2 of Revision 4, the event-specific methodology has been validated as appropriate for the NPM associated with the Standard Design Approval Application (SDAA) of the US460 design (i.e., the NPM-20) as follows:

The methodology used in Revision 3 was considered a starting point.

Nuscale performed analyses supporting other design work between DCA and SDAA. Prior to developing the SDAA for the US460, other potential designs were investigated. The preliminary analyses associated with this design work validated the continued applicability of the Revision 3 methodology or otherwise identified required changes. The changes were incorporated in Revision 4.

NuScale performed preliminary analyses supporting the SDAA. The preliminary analyses validated the continued applicability of the Revision 3 methodology or otherwise identified required changes. The changes were incorporated in Revision 4.

NuScale performed the final analysis supporting the Final Safety Analysis Report (FSAR) Revision 0. The final analyses validated the continued applicability of the Revision 3 methodology or otherwise identified required changes. The changes were incorporated in Revision 4.

NuScale updated the event-specific analyses to address NRC audit questions and RAIs received during the SDAA review, incorporate design changes made during the SDAA review, and resolve issues identified through the NuScale Corrective Action Program. The updated analyses are to be incorporated in FSAR Revision 2. The updated analyses validated the applicability of the methodology identified in Revision 4 or otherwise identified required changes. The changes were made as markups to Revision 4. Because of the process used above, the event-specific methodology in Section 7.2 of Revision 4, including markups, is applicable to the NPM-20 of the US460 design. For this reason, item i of part d of the NRC request is not applicable. Regarding item ii of part d, the technical justification is provided throughout TR-0516-49416, as described, and is confirmed by the event-specific analyses for NPM-20 supporting the FSAR. These analyses demonstrate NuScale Nonproprietary NuScale Nonproprietary

conservative results for NPM-20 using the methodology in Section 7.2 of Revision 4, including markups. The NRC was provided the event-specific analyses supporting the FSAR for audit as part of the SDAA review. The fact that NuScale has made updates to Section 7.2 in Revision 4 and in the markups is evidence that the methodology has been revised to be applicable to, and consistent with, the NPM-20 of the US460 design. Impact on Topical Report: Topical Report TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology, has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary NuScale Nonproprietary

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 4 NRELAP5 code was derived from the Idaho National Laboratory (INL) RELAP5-3D© computer code. RELAP5-3D©, version 4.1.3 was used as the baseline development platform for the NRELAP5 code. RELAP5-3D© was procured and commercial grade dedication was performed by NuScale. Subsequently, features were added and changes made to address unique aspects of an NPM design and licensing methodology. NRELAP5 is a non-homogenous, non-equilibrium two-fluid thermal hydraulic systems analysis code capable of performing non-LOCA system transient analyses for an NPM. The NRELAP5 code has a heat conduction and heat transfer package that is similar in capability to other thermal-hydraulic codes in its class (such as TRAC, RETRAN or TRACE). It includes the trips and logic control systems that enable simulation of the plant protection and control system logic for analysis of a non-LOCA event in an NPM. The NRELAP5 code is described in the NuScale LOCA Evaluation Model licensing topical report. The NRELAP5 code has been assessed against several separate effects and integral effects tests as part of the code development and development of the NuScale LOCA evaluation model to demonstrate the capability to simulate an NPM response to LOCA events. Phenomena identified as high-ranked for the non-LOCA transients were evaluated with respect to the high-ranked phenomena identified and assessed as part of the NuScale LOCA evaluation model development. Additional validation of NRELAP5 against separate effects testing, integral effects testing, and code to code benchmarking, were performed as necessary to justify applicability of the NRELAP5 code for non-LOCA system thermal-hydraulic analysis. High-ranked phenomena for non-LOCA events that were not assessed as part of the NuScale LOCA evaluation model development were therefore addressed in different ways:

1. additional NRELAP5 code assessment performed against separate effects or integral effects test data

2. code-to-code benchmark performed between NRELAP5 and independent system thermal-hydraulics code

3. phenomena addressed as part of the downstream subchannel analysis

4. phenomena addressed by specifying appropriately conservative input to the system transient analysis In particular, separate and integral effects testing were performed at the NIST-1 and NIST-2 facilities to support applicability of the NRELAP5 code for non-LOCA system transient analysis.

Separate effects testing of the DHRS was performed. Integral effects testing of an NPM response to a decrease in secondary side heat transfer, and integral effects testing of DHRS operation were performed. A code-to-code benchmark was performed to compare the NRELAP5 and RETRAN-3D responses to a range of reactivity insertion conditions in an NPM. Computational fluid dynamics was used to independently assess the heat transfer correlation used in NRELAP5 for the helical SGs of the NuScale designs. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The general non-LOCA transient analysis process is described in this report. The general methodology for conservatively biasing initial and boundary conditions for event analysis is presented. Each initiating event is then considered to identify the acceptance criteria that may be challenged during the event. For each non-LOCA event, a description of the event is provided including biases and conservatisms applied, sensitivity studies performed, single active failures and loss of power scenarios that challenge the event acceptance criteria. For each transient

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 5 event, the acceptance criteria where margin to the limit may be challenged are identified. For these acceptance criteria, sensitivity calculations are performed as necessary to confirm that appropriately conservative inputs are specified and to determine conditions that result in minimum margin. For other acceptance criteria where margin to the limit is not challenged, representative results from the overall scope of sensitivity calculations performed are sufficient to demonstrate that margin to the acceptance criterion is maintained. For non-LOCA initiating events that actuate the DHRS, the EM is applicable for the short-term transient progression; during this time frame the mixture level remains above the top of the riser and primary side natural circulation is maintained. For selected non-LOCA events, representative system transient results are provided to demonstrate the application of the evaluation model for an NPM. System transient calculations are executed for sufficient duration to demonstrate that the initiating event is mitigated and stable cooling is established. Results of representative calculations show that the maximum primary and secondary pressure acceptance criteria are not significantly challenged in an NPM design. Margin to other quantitative acceptance criteria for minimum critical heat flux ratio, fuel centerline temperature, and radiological dose limits applicable for the non-LOCA events are demonstrated as part of downstream subchannel or accident radiological analyses that are described in separate reports. In addition, long-term cooling analysis methodology is presented in a separate report. NuScale requests U.S. Nuclear Regulatory Commission (NRC) approval to use the EM described in this report for analyses of NPM design basis non-LOCA events that require system transient analysis. The specific scope of the non-LOCA events for which the EM applies is delineated in Section 1.2. NuScale is not seeking NRC approval of the representative calculations that are described in this report.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 32 4.2 Design Basis Event Acceptance Criteria Safety analyses are performed to demonstrate that a nuclear power plant can meet applicable acceptance criteria for a limiting set of AOOs, IEs, and accidents. If the risk of an event is defined as the product of the events frequency of occurrence and its consequences, then the design of a plant should be such that events produce about the same level of risk. The acceptance criteria indicated by the GDC for nuclear power plants (Reference 4) reflect the risk of an event. Relatively frequent events such as AOOs are prohibited from resulting in serious consequences, but relatively rare events (postulated accidents) are allowed to produce more severe consequences. Design basis events for an NPM are categorized as AOOs, IEs, or postulated accidents. Table 4-2, Table 4-3, and Table 4-4 summarize the acceptance criteria applied for AOOs, IEs, and postulated accidents, respectively. The applicable acceptance criteria identified for each event are based on the event classification as identified in Table 4-1. For a limited number of events, a more conservative acceptance criterion may be applied than required based on the event classification. For many non-LOCA transient events, the specific acceptance criterion is not challenged during the event progression. For example, events that result in an increase in heat removal from the RCS may have a maximum RCS pressure higher than the initial operating pressure, but does not challenge the margin to the maximum RCS pressure acceptance criterion. In contrast, events that result in a decrease in heat removal from the RCS may result in an RCS pressurization that could challenge the maximum RCS pressure acceptance criterion. In Section 7.2, the acceptance criteria of interest for each non-LOCA event are identified. The acceptance criteria of interest are those where margin to the limit may be challenged during the event progression. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 In the event-specific transient analysis, sensitivity calculations are performed as necessary to ensure that the event meets acceptance criteria that may be challenged. These sensitivity calculations are performed to confirm that appropriately conservative inputs are specified to identify the case that results in minimum margin to the acceptance criterion of interest. For other acceptance criteria where margin to the limit is not challenged, representative results from the overall scope of sensitivity calculations performed are sufficient to demonstrate that margin to the acceptance criterion is maintained. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 A prime example of an acceptance criterion where an NPM design has significant margin is the maximum secondary system pressure. Unlike in typical PWR designs, in a NuScale design, the design pressure of the SG secondary side up to the second containment isolation valves is equal to the RCS design pressure. This feature supports the design and operation of the SG and DHRS. In a non-LOCA event that results in DHRS actuation, typically the maximum secondary side pressure occurs in the first minutes of the transient progression, following DHRS actuation. After DHRS is actuated, the fluid in the DHRS flows into the SG. Heat is transferred from the RCS primary system to the SG, where the

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 33 DHRS loop inventory boils in the SG tubes. The steam flow is then condensed in the DHRS condensers and the energy is transferred to the reactor pool UHS. The maximum pressure in the SG secondary side is limited to the saturation pressure at the temperature of the RCS fluid on the SG primary side. Therefore, the maximum secondary pressure is affected by the secondary side inventory and the primary side conditions at the time of DHRS actuation, and is less sensitive to a specific initiating event. The SG design pressure is significantly higher than pressures expected during DHRS operation. The margin to the SG design pressure is physically limited, based on the primary side conditions. The representative transient results in Section 8.0 demonstrate that significant margin to the maximum SG pressure acceptance criterion is maintained for all types of events. Therefore, extensive sensitivity calculations to maximize secondary side pressure are not necessary for the non-LOCA transient analysis calculations. (( }}2(a),(c) Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Another example is RCS pressure. In the NPM, operation of an RSV at its lift setpoint is sufficient to mitigate an increasing pressure with minimal overshoot. This design feature ensures that the maximum RCS pressure does not significantly exceed the RSV lift setpoint. With appropriate selection of the RSV lift setpoint, including consideration of uncertainty, the maximum RCS pressure will be well below the acceptance criterion. Event analyses where pressure increases during the transient apply biases and include sensitivities to reach conditions where RSV lift may occur. If RSV lift occurs, modification of other biases can generate many cases with similar maximum RCS pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases with RSV lift. If RSV lift does not occur for an event, then RCS pressure will not challenge the acceptance criterion and extensive sensitivity studies are not required to investigate differences in RCS pressure below the RSV lift setpoint. (( RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 40 4.3.1.1.2 Axial Power Shape For the system transient analysis, a single channel core model is used, as described in Section 6.0. A nominal center-peaked average axial power shape is input for the single core channel for consistency with development of the reactivity feedback coefficients determined in the core design analyses. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Uncertainties associated with the axial power shape and axial and radial power peaking factors that can affect the minimum critical heat flux ratio (MCHFR) and peak centerline fuel temperature are accounted for in the downstream subchannel analyses as described in Reference 6, supplemented by Reference 28. The design features of an NPM preclude challenge to the primary and secondary pressure acceptance criteria as discussed in Section 7.0 and as shown in the representative results in Section 8.0. Sensitivity studies on the axial power shape confirm that the primary and secondary system pressure, flow and fluid temperature responses are not significantly affected by the axial power shape. Therefore, use of a nominal center-peaked average axial power shape input is appropriate for the system transient analyses. 4.3.1.1.3 Energy Deposition Factor The energy deposition factor is the portion of the energy generated in the core that is directly deposited in the fuel. A bounding high energy deposition factor that results in all energy being deposited in the fuel is used in the non-LOCA analyses. For cooldown or reactivity insertion events that cause total core power changes, increasing the core energy deposited in the fuel slows the thermal-hydraulic response of the system to changing power levels. The effect of the energy deposition factor on the primary system pressure response is generally insignificant except in very fast reactivity events such as control rod ejection where Doppler reactivity feedback is important; control rod ejection analysis is outside the scope of this EM. Sensitivity studies on this parameter confirm that its effect on the system response is not significant with respect to demonstrating margin to acceptance criteria for NPM events addressed by the non-LOCA evaluation model due to the MPS design and selection of the analytical limits to actuate reactor trip. 4.3.1.2 Interface with Fuel Rod Performance Design (Input to the Transient Analysis) Fuel rod design analysis performed using a fuel performance code approved for a NuScale design provides input to the system transient analysis. The NuScale transient analysis methodology using NRELAP5 can be applied to a typical light water reactor fuel assembly design, and does not require that a specific code or suite of codes be used for the fuel performance analysis. The fuel assembly geometry, required material properties, and the fuel performance data needed for

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 532 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For each event analyzed following the non-LOCA evaluation model, a description of the event progression, significant inputs and results, and representative results of sensitivity studies are presented in the following sections. Sensitivity studies are performed to identify plant conditions that result in bounding transient analyses. Studies that identify acceptance criteria challenges or bounding transient forcing functions are discussed as well. Other sensitivity studies that determine bounding inputs for RCS flow, fuel parameters, etc. may not necessarily be discussed for every event. The selection of parameters to be studied is focused on the acceptance criteria challenged by the event. For events where a parameter has minimal or no impact on the consequences, the bias or conservatism identified for the parameter is that typically applied; use of an alternate assumption is also acceptable. The sensitivity study results provided herein are for a representative NPM (NPM-160). Sensitivity studies are repeated as necessary for different NPM designsperformed as part of the implementation of the methodology are identified in the following sections. In addition, Section 7.2.1.3 provides example sensitivity studies for one specific event that are then used to generally define and demonstrate the methodology for the events in Section 7.2. Initial RCS flow is typically biased to the low condition in all event simulations because this is bounding for MCHFR. (( }}2(a),(c) Steam generator tube plugging is considered for each event in the "Initial conditions, biases, and conservatisms" tables. The term, "Biased to the low condition" indicates no tube plugging is assumed. Biased to the high condition indicates (( }}2(a),(c) steam generator tube plugging. In an NPM design, the rod control system is set to insert only mode at full power to prevent automatic withdrawal of the control bank at full power. Although this plant feature exists, the feature is not credited during events where control rod withdrawal results in a bounding result. Containment integrity(2) < Limits < Limits < Limits Escalation of an AOO to an accident (AOO) or consequential loss of system functionality (IE or accident)? No No No Dose(1) Normal operations < Limit < Limit

1. This criterion is confirmed as part of a separate follow-on analysis.

2. Containment integrity is evaluated by a separate analysis methodology.

3. If the minimum CHFR is less than or equal to the CHFR analysis limit, or if the maximum fuel centerline temperature exceeds the melting temperature, the fuel rod is assumed to be failed. If fuel failure is calculated, this is accounted for in the downstream radiological dose analysis.

Table 7-4 Regulatory acceptance criteria (Continued) Description AOO Criteria IE Criteria Accident Criteria

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 535 conservative alternative to accounting for the decalibration effect on that power-related reactor trip signal in the analysis. To maximize the overall feedwater temperature change, the feedwater temperature transient starts at the initial (full power) feedwater temperature biased to the high condition, and terminates at the coldest temperature in the secondary, which is saturation temperature at condenser vacuum conditions. A sensitivity study on feedwater temperature cooldown rate is performed to identify the rate that results in limiting conditions. (( }}2(a),(c) RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Additional sensitivity studies are performed on other parameters, as identified in Table 7-7necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation, covered by a separate methodology. Table 7-5 Acceptance criteria, single active failure, loss of power scenarios - decrease in feedwater temperature Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion MCHFR CHF is the challenged acceptance criterion for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) No single failure The challenging cases typically occur when all equipment operates as designed. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 537 7.2.1.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-7 are considered in identifying a bounding transient simulation for MCHFR.53 Audit Question A-NonLOCA.LTR-28 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition Initial PZR level Varied.Biased to the high condition. Initial feedwater temperature Varied. Initial fuel temperature Nominal. Moderator Temperature Coefficient (MTC) Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal RSV lift setpoint Nominal }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 538 SG tube plugging Biased to the low condition. Minimum feedwater temperature Biased to the low condition. Feedwater temperature cooldown rate Spectrum.Varied RCS Temperature Control Automatic rod control Varied. Boron concentration Not credited. PZR Pressure Control PZR spray Varied.Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 539 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. For example, sensitivity studies are performed to consider the effects of fuel-related parameters (initial fuel temperature, time in life). Representative results for these studies are presented in Table 7-8 and Table 7-9. (( }}2(a),(c) Steam Pressure Control Turbine throttle valves Varied Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c) Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 540 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The next example sensitivity study, Table 7-10, identifies the feedwater temperature decrease rate that minimizes MCHFR. (( }}2(a),(c) RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The last example sensitivity study for this event evaluates the effects of steam system boundary condition type on MCHFR and assesses the effects of a single failure of an MSIV to isolate. Representative results are presented in Table 7-11. The boundary condition in Table 7-11 is a constant steam flow, which allows the steam pressure to move in response to the transient. The previous study in Table 7-10 models a constant steam pressure boundary condition that allows flow to vary while maintaining pressure. (( }}2(a),(c) RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The results of these studies are provided to demonstrate the process by which impact of the the bias of a parameter can be determined. For sensitivity studies that show a response similar to Table 7-8, the limiting bias should be selected. For sensitivity studies that show a response similar to Table 7-9, there is not a bias direction that is limiting and the parameter can be selected based on other analysis considerations. Other events in Section 7.2 use the same general sensitivity study process. Table 7-8 Representative fuel exposure study (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 541 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-9 Representative fuel temperature study (( }}2(a),(c) Table 7-10 Representative feedwater temperature transient studyNot Used (( }}2(a),(c) Table 7-11 Representative boundary condition type / single active failure studiesNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 542 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The results of these representative sensitivity studies are provided as examples for this event. However, the sensitivity studies also demonstrate the methodology used generally for the events in Section 7.2.

Table 7-8 demonstrates how to vary a parameter that has a range of possible initial values. (( }}2(a),(c)

Table 7-9 also demonstrates how to vary a parameter that has a range of possible initial values. (( }}2(a),(c)

Table 7-10 demonstrates how to consider a spectrum of values for a parameter. (( }}2(a),(c)

Table 7-11 demonstrates how to vary a control system parameter that has two basic states (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 543 (( }}2(a),(c)

Based on the results of sensitivity studies for individual parameters, engineering judgment is used to determine whether a figure of merit is sensitive to a specific parameter. The parameters that show sensitivity in the figure of merit are then combined to identify limiting cases for the figure of merit. (( }}2(a),(c) No specific quantification is provided to define the threshold for sensitive vs. insensitive. Instead, a two-tiered approach is used based on the available margin for the figure of merit as described in Section 4.2. When the margin to an acceptance criterion is less than the thresholds defined in Section 4.2, a reduction in margin due to change in a parameter should not be considered negligible. 7.2.2 Increase in Feedwater Flow The methodology used to simulate a postulated increase in feedwater flow for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.2.1 General Event Description A feedwater system malfunction that causes an increase in feedwater flow results in an unplanned overcooling of the RCS. The subsequent decrease in RCS temperature increases core reactivity due to moderator feedback, which raises reactor power. Decreasing average RCS temperature also prompts the control rod controller to withdraw the regulating bank from the core if automatic control is enabled. Rising reactor power can cause RTS actuation on the high power or high pressurizer pressure signal. The feedwater flow increase can also cause RTS actuation on low steam line superheat or high steam line pressure. DHRS also actuates on one of the several signals that cause RTS. Closure of the FWIVs following DHRS actuation isolates the SGs from the feedwater source, ending the overcooling event. Core decay heat drives natural circulation, which transfers thermal energy from the RCS to the reactor pool via the DHRS. Passive DHRS cooling is established and the transient calculation is

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 545 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 (( }}2(a),(c). Additional sensitivity studies are performed on other parameters, as necessaryidentified in Table 7-14, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. Audit Question A-NonLOCA.LTR-60 7.2.2.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-13. Table 7-12 Acceptance criteria, single active failure, loss of power scenarios - increase in feedwater flow Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion MCHFR SG overfill CHF is challenged for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) An increase in feedwater flow could cause the SG to overfill. No single failure Failure of one FWIV to close TExcept when considering SG overfill, the challenging cases typically occur when all equipment operates as designed. FWIV single failure is typically limiting for SG overfill. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event and the potential for SG overfill.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 547 Audit Questions A-NonLOCA.LTR-28, A-NonLOCA.LTR-65 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level VariedBiased to the high condition. Initial feedwater temperature Varied. Initial fuel temperature Nominal. MTC Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the low condition. }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 549 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, two sensitivity studies are performed to identify the most challenging secondary biases for the feedwater flow transient for MCHFR. (( }}2(a),(c) Steam Pressure Control Turbine throttle valves Enabled to control to constant steam pressure. Turbine bypass valves Disabled. Feedwater and Turbine Load Control fFeedwater pump speed Spectrum.N/A. CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c) Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 550 (( }}2(a),(c) Representative results for these studies are presented in Table 7-16. These results demonstrate an example of (( }}2(a),(c) as described in the general methodology discussion in Section 7.2.1.3. Audit Question A-NonLOCA.LTR-60 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The increase in feedwater flow also creates the potential for overfilling the SG. The assumptions in Table 7-12 and Table 7-14 related to identifying the limiting MCHFR do not necessarily result in the maximum SG level. The biases and conservatisms indicated in Table 7-15, and the single failure assumption in Table 7-12, are considered in identifying a bounding transient simulation for SG overfill.For example, a single failure of a feedwater isolation valve to close is expected to maximize SG level. Additional sSensitivity studies are performed on the Table 7-14 parameters, as necessary, to identify the case(s) with a potentially limiting SG level.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 551 Audit Question A-NonLOCA.LTR-60 RAI 10297 Question NonLOCA.LTR-60, RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-15 Initial conditions, biases, and conservatisms for SG overfill - increase in feedwater flow Not Used Parameter Bias / Conservatism Basis Initial reactor power Varied. (( }}2(a),(c) Initial RCS average temperature Varied. (( }}2(a),(c) Initial RCS flow rate Varied. (( }}2(a),(c) Initial PZR pressure Varied. (( }}2(a),(c) Initial PZR level Varied. (( }}2(a),(c) Initial feedwater temperature Varied. (( }}2(a),(c) Initial fuel temperature Biased to the high condition. (( }}2(a),(c) MTC Varied. (( }}2(a),(c) Kinetics Varied. (( }}2(a),(c) Decay heat Biased to the high condition. (( }}2(a),(c) Initial SG pressure(1) Biased to the low condition. (( }}2(a),(c) SG heat transfer Varied. (( }}2(a),(c) RSV lift setpoint Nominal. (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 552 SG tube plugging Varied. (( }}2(a),(c) RCS Temperature Control Automatic rod control Boron concentration Varied. Not credited. (( }}2(a),(c) PZR Pressure Control PZR spray PZR heaters Disabled. Nominal. (( }}2(a),(c) PZR Level Control Charging Letdown Not credited. Disabled. (( }}2(a),(c) Table 7-15 Initial conditions, biases, and conservatisms for SG overfill - increase in feedwater flow (Continued) Not Used Parameter Bias / Conservatism Basis

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 554 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.3 Increase in Steam Flow The methodology used to simulate a postulated increase in steam flow for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.3.1 General Event Description In an NPM design, two main steam lines exit the CNV, and combine to form a common main steam line that connects to the high pressure turbine. The turbine bypass and main steam safety valves are located in the common main steam line, which is downstream of the MSIVs. A spurious opening of the turbine bypass valve or a main steam safety valve would cause an increase in steam flow, which results in an unplanned overcooling of the RCS. The subsequent decrease in RCS temperature increases core reactivity due to moderator feedback, which raises reactor power. Decreasing average RCS temperature also prompts the control rod controller to withdraw the regulating bank from the core if automatic control is enabled (rod withdrawal at 100 percent power is inhibited, however, it is conservatively allowed in the analysis). Rising reactor power typically causes RTS actuation on a high power or high power rate signal. Additionally, a large increase in steam flow would rapidly depressurize the secondary system and could cause Table 7-16 Representative increase in feedwater flow study - high and low SG performance with maximum power and minimum RCS flowNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 556 (( }}2(a),(c) Neglecting reactor trip on a power-related reactor trip signal is a conservative alternative to accounting for the decalibration effect on that power-related reactor trip signal in the analysis. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The increase in steam flow event starts at the initial (full power) steam flow. Sensitivity studies are performed on the degree of steam flow increase, (( }}2(a),(c) to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. 7.2.3.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-18. Table 7-17 Acceptance criteria, single active failure, loss of power scenarios - increase in steam flow Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion MCHFR CHF is challenged for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) No single failure Note that a single active failure of a FWIV to close would occur after RTS and DHRS actuation, subsequent to when the MCHFR occurs. Consequently, the MCHFR occurs before the single active failure of an FWIV to close could affect the transient. Otherwise, the challenging cases typically occur when all equipment operates as designed. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event. Table 7-18 Acceptance criteria - increase in steam flow Acceptance Criteria Discussion Primary pressure Primary pressure initially drops as inventory shrinks due to increased heat removal. As reactor power increases and as the PZR heaters respond, an increase (typically less than 100 psi) in primary pressure is observed.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 558 Audit Question A-NonLOCA.LTR-28 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level Varied.Biased to the high condition. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Biased to the high condition. SG heat transfer Nominal RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Steam flow increase Spectrum.Varied }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 560 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, two sensitivity studies are performed to identify the most challenging increase in steam flow for MCHFR. (( }}2(a),(c) Representative results for these studies are presented in Table 7-20 and Table 7-21. These results provide an example of varying a parameter (( }}2(a),(c) as described in the general methodology discussion in Section 7.2.1.3. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c)

2. ((

}}2(a),(c) Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 561 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.4 Steam System Piping Failure Inside or Outside of Containment The methodology used to simulate a postulated steam system piping failure for an NPM, and an evaluation of the acceptance criteria listed in Table 7-4, are presented below. Since both split breaks (relatively higher event frequency) and double-ended guillotine breaks (relatively lower event frequency) are analyzed, the more restrictive AOO criteria for system pressures, critical heat flux ratio, and fuel centerline melt applicable to breaks with higher event frequency are used in the evaluation. Radiological dose consequences are assessed as part of the downstream accident radiological dose analysis, documented in a separate report, and compared against the appropriate acceptance criteria. Table 7-20 Representative steam flow study - nominal steam generator heat transferNot Used (( }}2(a),(c) Table 7-21 Representative steam flow study - steam generator heat transfer biased lowNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 564 7.2.4.2 Acceptance Criteria Evaluation of the most challenging case relative to the acceptance criteria is presented in Table 7-23. 7.2.4.3 Biases, Conservatisms, and Sensitivity Studies RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The biases and conservatisms indicated in Table 7-24, and a spectrum of break locations and sizes, are considered in identifying a bounding transient simulation for MCHFR and mass release. In the radiological analysis, primary-to-secondary leakage is assumed, which allows primary coolant to be released with the SG break flow. The transient SG mass release can be calculated for use as an input to the downstream radiological analysis. Alternatively, bounding assumptions for primary coolant release can be used in the radiological analysis to eliminate the need for calculating SG mass release. If such bounding assumptions are used, transient analysis to maximize SG mass release is not required. Table 7-23 Acceptance criteria - steam line break Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of this cooldown event, primary pressure remains below the acceptance criterion for peak primary pressure. Secondary pressure Due to the depressurizing nature of this cooldown event, secondary pressure remains below the acceptance criterion for peak secondary pressure. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Sensitivity cases are performed to support the follow-on MCHFR evaluation. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation to a more serious accident or consequential loss of functionality This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 567 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response and challenging mass releases for this overcooling event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Representative sensitivity study results for MCHFR are presented in Table 7-25. In the cases modeling breaks outside of containment, the MSIV on the train with the break is assumed to fail to close, resulting in the complete emptying of the Steam Pressure Control Turbine throttle valves EnabledVaried. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled. (( }}2(a),(c) Table 7-24 Initial conditions, biases, and conservatisms - steam line break (Continued) Parameter Bias / Conservatism Basis(1) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 568 affected SG. In cases modeling breaks inside of containment, the FWIV on the train with the break is assumed to fail to close. Break size is the faction (percent) of the pipe cross-sectional area. (( }}2(a),(c) The sensitivity study demonstrates using a spectrum analysis (( }}2(a),(c) consistent with the general methodology discussion in Section 7.2.1.3. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-25 Representative steam line break studyNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 571 7.2.5.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-28 are considered in order to identify the bounding transient simulation for MCHFR for the CNV flooding/loss of CNV vacuum event. Audit Question A-NonLOCA.LTR-28 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-27 Acceptance criteria - containment flooding / loss of containment vacuum Acceptance Criteria Discussion Primary pressure Due to the depressurizing nature of the event, sensitivities that maximize primary pressure are not analyzed. Peak primary pressure resulting from CNV flooding/loss of CNV vacuum is bounded by other AOO events. Secondary pressure Due to the primary system depressurization of this cooldown event, secondary pressure remains below the acceptance criterion for peak secondary pressure. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 572 Initial PZR level Nominal Initial feedwater temperature Nominal Initial fuel temperature Nominal MTC Biased to the EOC condition. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal SG heat transfer Nominal RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Initial containment pressure Varied Initial pool temperature Varied RCCW leak flow Varied RCCW temperature Varied RCS Temperature Control Automatic rod control Enabled. Boron concentration Not credited. Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 573 PZR Pressure Control PZR spray Varied.Disabled. PZR heaters EnabledNominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 574 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response for this overcooling event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 As an example, the initial reactor power, pressurizer pressure, RCS average coolant temperature, RCS flow rate, containment pressure, pool temperature, RCCW leak flow rate, RCCW temperature arewere varied for a representative NPM for the CNV flooding/loss of CNV vacuum event as indicated in Table 7-29. The sensitivity studies indicated that the CNV flooding cases are more challenging to MCHFR than loss of CNV vacuum. The sensitivity study results also indicated that for a variety of initial RCS conditions, reactor pool conditions, and condition of liquid or air ingress to containment, loss of containment vacuum or containment flooding results in a slow overcooling transient that is non-limiting with respect to MCHFR compared to other cooldown event AOOs. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled. (CNV flooding) N/A. (CNV vacuum loss)

1. ((

}}2(a),(c) Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 575 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.6 Turbine Trip / Loss of External Load The methodology used to simulate a postulated turbine trip/loss of external load (LOEL) for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.6.1 General Event Description Turbine trip initiates with a turbine stop valve closure while loss of external load initiates with a turbine control valve closure. Otherwise, these transients are essentially equivalent and result in the sudden removal of the secondary side heat sink, overpressurization of the secondary system, and overheating of the RCS. Rising system pressures typically result in RTS actuation on the high PZR or steam pressure signal. Reactor trip and transition to stable DHRS flow terminates Table 7-29 Representative sensitivity studies - containment flooding / loss of containment vacuumNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 576 the transient with the NPM in a safe, stable condition. Table 7-30 lists the relevant acceptance criteria, SAF, and LOP scenarios. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The limiting pressure responses typically occur when the event is initiated from full power conditions, and the initial conditions are biased in the conservative directions. 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. 7.2.6.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-31. Table 7-30 Acceptance criteria, single active failure, loss of power scenarios - turbine trip / loss of external load Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion Primary pressure, secondary pressure Primary and secondary pressures are challenged during this overheating event. Failure of one FWIV to close Typically challenging for secondary side pressurization cases (negligible effect for primary side pressurization cases, which assume loss of AC power at event initiation and therefore feedwater is lost). Loss of AC power at transient initiation No loss of power Typically maximizes primary pressure (feedwater is lost). Typically maximizes secondary pressure. Table 7-31 Acceptance criteria - turbine trip / loss of external load Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation and continued FW operation, in addition to the actual turbine trip or loss of external load. The DHRS heat removal is limited by the DHRS condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 579 Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. The maximum RCS pressure is limited by the RSV lifting, with its applied lift pressure bias, and therefore many cases may have similar peak pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity study results are presented in Table 7-33. The results indicate that maximum RPV pressure occurs when PZR pressure is biased to the high condition, average temperature is biased to the low condition, and a loss of power leads to a loss of the secondary side heat sink (since feedwater is lost, the failure of an FWIV to close would have no effect on the event). The results also confirm that many cases have similar maximum RPV pressures and that there is significant margin to the acceptance criterion (2310 psia for this representative NPM). With respect to secondary side SG pressure, maximum SG pressure occurs when average temperature and SG pressure are biased to the high condition, and all power sources are available - the failure of an FWIV to close increases the peak SG pressure by a very small amount. The results also confirm that many cases have similar maximum secondary side pressures and remain well below the secondary design pressure (2100 psia for this representative NPM). Steam Pressure Control Turbine throttle valves Disabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

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}}2(a),(c) Table 7-32 Initial conditions, biases, and conservatisms - turbine trip / loss of external load (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 580 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.7 Loss of Condenser Vacuum The methodology used to simulate a postulated loss of condenser vacuum for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4 are presented below. 7.2.7.1 General Event Description Loss of condenser vacuum initiates with a turbine stop valve closure. Also, a loss of condenser vacuum is postulated to lead to a loss of feedwater flow. Turbine trip and loss of feedwater result in the sudden removal of the secondary side heat sink, pressurization of the secondary system, and overheating of the RCS. Rising system pressures typically result in a rapid RTS actuation on either high PZR or steam pressure. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. The loss of condenser vacuum event is essentially equivalent to the turbine trip / loss of external load events discussed in Section 7.2.6. The main difference is that the loss of condenser vacuum event includes a loss of feedwater flow at event initiation. However, because the turbine trip / loss of external load events consider a loss of normal AC power at event initiation, those events also model a loss of feedwater flow at event initiation. As a result, the scenarios analyzed as part of Section 7.2.6 address the loss of condenser vacuum event. Therefore, the relevant acceptance criteria, SAF, and LOP scenarios from Table 7-30 are also applicable to the loss of condenser vacuum event. Table 7-33 Representative sensitivity studies - turbine trip / loss of external loadNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 582 The MSIV closure event can occur when one or both MSIVs close unexpectedly. The limiting pressure responses typically occur when the event is initiated from full power conditions, and the initial conditions are biased in the conservative directions. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies on number of MSIVs closing, 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 identified in Table 7-40necessary, to identify the case(s) with the potentially limiting peak primary and secondary pressures. 7.2.8.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-39. Table 7-38 Acceptance criteria, single active failure, loss of power scenarios - main steam isolation valve closure Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion Primary pressure, secondary pressure Primary and secondary pressures are challenged during this overheating event. No single failure Failure of one FWIV to close Typically challenging for primary pressure. Typically challenging for steam generator pressure. Loss of AC Power at transient initiation No loss of power Typically maximizes primary pressure. Typically maximizes secondary pressure. Table 7-39 Acceptance criteria - main steam isolation valve closure Acceptance Criteria Discussion Primary pressure Primary pressure quickly rises to the peak value, then drops as the lowest setpoint RSV lifts to reduce pressure. Secondary pressure Peak secondary pressurization is largely a function of DHRS actuation and continued FW operation, in addition to the actual main steam isolation valve closure. The DHRS heat removal is limited by the DHRS condenser so some pressurization is expected for every actuation of this system. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 585 Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. The maximum RCS pressure is limited by the RSV lifting, with its applied lift pressure bias, and therefore many cases may have similar peak pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity study results are presented in Table 7-41. The results presented consider the closure of two MSIVs, which isolates both steam generators, and no single active failure. Maximum pressures are similar for cases in which the RSV actuation limits the RPV pressurization and remain well below the acceptance criterion (2310 psia for this representative NPM). For example, the maximum calculated pressure for the peak RPV pressure study is approximately equal for two additional cases, namely, decreased steam generator heat transfer and decreased feedwater temperature. Additionally, the results of the sensitivity studies indicate that maximum RPV pressure occurs when PZR pressure is biased to the high condition, RCS flow is at the nominal value (if maximum RCS pressure is below the RSV lift pressure), and a loss of AC power leads to a loss of the secondary side heat sink (loss of AC power occurs coincident with reactor trip). With respect to secondary side SG pressure, maximum SG pressure occurs when average temperature is biased to the high condition, RCS flow is biased to the low condition, and all power sources are available. The results also confirm that many cases have similar maximum secondary side pressures and remain well below the secondary design pressure (2100 psia for this representative NPM). feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

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}}2(a),(c) Table 7-40 Initial conditions, biases, and conservatisms - main steam isolation valve closure (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 586 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.9 Loss of Nonemergency AC Power The methodology used to simulate a postulated loss of nonemergency (normal) AC power for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.9.1 General Event Description The low voltage AC electrical distribution system (ELVS) supplies AC power to plant motors, heaters, packaged equipment, and battery chargers. Loss of normal AC power to the station auxiliaries can result from electrical grid-related failures, failures in plant or switchyard equipment, or external weather events. The nonsafety-related EDNS and EDAS/EDSS may remain available via battery operation; the primary loads for these systems include the module control system (MCS) and the MPS. Loss of the EDNS and/or EDAS/EDSS batteries with the loss of normal AC power is considered as described in Section 7.1.3. A loss of AC power results in a pressurization of the secondary system and overheating of the RCS. Reactor trip and transition to stable DHRS flow terminates the transient with the NPM in a safe, stable condition. Table 7-42 lists the relevant acceptance criteria, SAF, and LOP scenarios. Table 7-41 Representative sensitivity studies - main steam isolation valve closureNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 587 The electrical system design may vary by NPM design. Review of the electrical system is performed to determine the impact on plant equipment from the loss of power. In general, the following typically occur from the loss of power (AC or DC or both):

turbine trip occurs

the feedwater pumps and CVCS pumps stop

the PZR heaters turn off

control rods begin to drop due to loss of power to control rod drive mechanisms (CRDMs)

the MPS actuates reactor trip, DHRS, and various system isolations within 60 seconds after event initiation (if not already actuated during that time), due to loss of AC power to the EDAS/EDSS battery chargers

the MPS actuates ECCS within 24 hours after event initiation, due to loss of AC power to the EDAS/EDSS battery chargers RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Of the various scenarios, the limiting one is typically the scenario that does not result in an immediate reactor trip or full control rod insertion at event initiation. Typically this scenario is the loss of the ELVS at event initiation, with EDNS and EDAS/EDSS available. Cases considering the loss of EDSS battery backup coincident with the initiating event are considered as described in Section 7.1.3. Consequently, the limiting pressure responses typically occur when the event is initiated from full power conditions, reactor trip is delayed until MPS actuation, and the initial conditions are biased in the conservative directions. 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 necessaryidentified in Table 7-44, to identify the case(s) with the potentially limiting peak primary and secondary pressures. Table 7-42 Acceptance criteria, single active failure, loss of power scenarios - loss of normal AC power Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion Primary pressure, secondary pressure Primary and secondary pressures are challenged during this overheating event. No single failure The challenging cases for primary pressure typically occur when all equipment is operational. Since feedwater is lost at transient initiation, peak secondary pressures are insensitive to the single failure of an FWIV to isolate. Loss of AC power at transient initiation Initiating event.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 590 Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. The maximum RCS pressure is limited by the RSV lifting, with its applied lift pressure bias, and therefore many cases may have similar peak pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies are performed to identify the highest primary and secondary side pressures and presented in Table 7-45. The Charging Not credited. Letdown Enabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

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}}2(a),(c)

2. Loss of normal AC power initiating event results in a loss of system function by loss of power to the system thereby making the system control not relevant to the event.

Table 7-44 Initial conditions, biases, and conservatisms - loss of normal AC power (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 591 results of the sensitivity studies indicate that the lowest-setpoint RSV operates to mitigate peak primary side pressurization (( }}2(a),(c) The peak secondary side pressures remain well below the design pressure (2100 for this representative NPM). RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-45 Representative sensitivity studies - loss of normal AC powerNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 593 7.2.10.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-48 are considered in identifying a bounding transient simulation for primary and steam generator pressure. Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Varied. Initial RCS flow rate Varied. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Varied. Initial fuel temperature Biased to the high condition. MTC Consistent with BOC kinetics. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal. }}2(a),(c) Table 7-47 Acceptance criteria - loss of normal feedwater flow (Continued) Acceptance Criteria Discussion

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 594 RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Feedwater flow decrease VariedSpectrum. RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed N/A. Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 595 Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the highest primary and secondary pressures, varying the magnitude of the feedwater flow rate decrease (other parameters are biased as indicated in Table 7-48). The maximum RCS pressure is limited by the RSV lifting, with its applied lift pressure bias, and therefore many cases may have similar peak pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative results of a sensitivity study for a spectrum of feedwater flow rate decreases are presented in Table 7-49. The results of the sensitivity study indicates that peak primary pressure occurs during a total loss of feedwater. Because the pressurization associated with the total loss of feedwater is sufficient to actuate the lowest-setpoint RSV, (( }}2(a),(c) well below the acceptance criterion of 2310 for this representative NPM). With respect to maximum secondary side SG pressure, maximum SG pressure occurs during a partial loss of feedwater, as the mismatch between primary heat production and secondary heat sink causes the RCS temperature to increase, and eventually actuates the RTS on high RCS riser temperature. The peak secondary side pressures remain well below the design pressure (2100 for this representative NPM). RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 CNV Pressure Control CNV evacuation system Disabled.

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}}2(a),(c) Table 7-49 Representative sensitivity studies - loss of normal feedwater flowNot Used (( }}2(a),(c) Table 7-48 Initial conditions, biases, and conservatisms - loss of normal feedwater flow (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 596 7.2.11 Inadvertent Decay Heat Removal System Actuation The methodology used to simulate a postulated inadvertent DHRS actuation for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. This event is unique to the NPM designs. 7.2.11.1 General Event Description Inadvertent actuation of the DHRS may result from either an unexpected DHRS valve actuation or a spurious DHRS actuation signal. Due to the DHRS design and configuration, several scenarios exist for consideration. The specific scenarios considered are discussed below. The sequence of events and the MPS signals that are reached first depend on the scenario. Reactor trip and transition to stable DHRS flow terminates the transients with the NPM in a safe, stable condition. The relevant acceptance criteria, SAF, and LOP scenarios are listed in Table 7-50. Scenario 1: The unexpected opening of a single DHRS valve can occur at full power or reduced power conditions. At low power conditions, a portion of the DHRS liquid inventory drains into the feedwater line, which momentarily increases feedwater flow and causes overcooling. This overcooling event is not considered further because it is bounded by other, more limiting overcooling events (i.e., increase in feedwater flow). The most challenging conditions for this heatup scenario occur at full power with initial conditions biased in the conservative directions. Since feedwater flow tends to increase in response to the reduced steam enthalpy and turbine load, limiting the feedwater response maximizes the heatup. Scenario 2: An inadvertent actuation signal isolates one steam generator, and initiates one DHRS train. This scenario is typically bounded by Scenario 3 discussed below. Table 7-49 Representative sensitivity studies - loss of normal feedwater flowNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 597 Scenario 3: An inadvertent actuation signal isolates both steam generators, and initiates both DHRS trains. This scenario represents a complete loss of normal heat removal from the RCS. The limiting conditions for this heatup scenario occur at full power with initial conditions biased in the conservative directions. Scenario 4: An inadvertent isolation of one steam generator. The valves associated with one SG are closed, but the associated DHRS train is not actuated. The heatup caused by the isolation of one SG causes an increase in primary pressure that results in reactor trip and RSV opening. The actuation of DHRS occurs later in the transient after MPS signals are reached. Secondary pressure peaks after DHRS actuation. Scenario 5: An inadvertent isolation of both steam generators. The valves associated with both SGs are closed, but neither DHRS train is actuated. The response is similar to Scenario 4, except the increase in primary pressure is more rapid and results in earlier reactor trip and RSV opening. The actuation of DHRS occurs later in the transient after MPS signals are reached. Secondary pressure peaks after DHRS actuation. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Each scenario is considered to determine the limiting cases for the acceptance criteria. Sensitivity studies on initial primary and secondary conditions are performed as identified in Table 7-52needed to identify the conditions that maximize peak system pressures. 7.2.11.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-51. Table 7-50 Acceptance criteria, single active failure, loss of power scenarios - inadvertent decay heat removal system actuation Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion Primary pressure, secondary pressure Challenged during the heatup scenarios of this event. No single failure Failure of one FWIV to close The challenging primary pressurization cases typically occur when all equipment is operational. Potentially challenging for secondary pressure, but typically bounded by other heatup events. No loss of power Maximizes system pressures.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 600 Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this event. The maximum RCS pressure is limited by the RSV lifting, with its applied lift pressure bias, and therefore many cases may have similar peak pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c) Table 7-52 Initial conditions, biases, and conservatisms - inadvertent decay heat removal system actuation (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 601 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies performed to identify the highest primary and secondary side pressure, varying primary conditions, steam generator heat transfer and pressure, are presented in Table 7-53. Unless otherwise noted, cases use BOC kinetics. The results of the sensitivity studies indicate that peak primary pressure occurs when both trains of DHRS inadvertently actuate (i.e., total loss of normal heat sink). The lowest-setpoint RSV operates to mitigate peak pressurization well below the acceptance criterion of 2310 for this representative NPM). (( }}2(a),(c) The peak secondary side pressures remain well below the design pressure (2100 for this representative NPM). RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-53 Representative sensitivity studies - inadvertent decay heat removal system actuationNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 602 7.2.12 Feedwater System Pipe Break Inside or Outside Containment The methodology used to simulate a postulated feedwater system pipe break for an NPM, and an evaluation of the acceptance criteria listed in Table 7-4, are presented below. Since both split breaks (relatively higher event frequency) and double-ended guillotine breaks (relatively lower event frequency) are analyzed, the more restrictive AOO criteria for system pressures, critical heat flux ratio, and fuel centerline melt applicable to breaks with higher event frequency are used in the evaluation. 7.2.12.1 General Event Description A feedwater line break can occur inside or outside of containment, and can range in size from a small split crack to a double ended rupture. A large feedwater line break inside containment results in a loss of containment vacuum and a high containment pressure MPS signal that actuates a reactor trip, isolates the secondary system and CVCS, and opens the DHRS valves. The steam generator, DHRS piping, and DHRS condenser on the affected side drain through the break. The non-affected steam generator system and DHRS loop provide cooling to the RCS via heat transfer to the reactor pool. The response of smaller feedwater line breaks inside containment is similar except that other MPS setpoints, such as high PZR pressure, may be reached before high containment pressure. A feedwater line break outside containment causes a loss of feedwater flow to the steam generators and a heatup of the RCS. Larger breaks result in rapid heatup events that pressurize the RCS beyond the high PZR pressure analytical limit. Smaller breaks cause a more gradual heatup, loss of secondary pressure, and Table 7-53 Representative sensitivity studies - inadvertent decay heat removal system actuationNot Used (Continued) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 604 7.2.12.3 Biases, Conservatisms, and Sensitivity Studies RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The biases and conservatisms presented in Table 7-56, and a spectrum of break locations and sizes, are considered in identifying the bounding transient simulation for primary and steam generator pressure. Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation to a more serious accident or consequential loss of functionality This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied. Initial PZR level Varied. }}2(a),(c) Table 7-55 Acceptance criteria - feedwater line break (Continued) Acceptance Criteria Discussion

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 605 Initial feedwater temperature Varied. Initial fuel temperature Biased to the high condition. MTC Consistent with BOC kinetics. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Nominal. Break size / location VariedSpectrum. RCS Temperature Control Automatic rod control Disabled. Boron concentration Not credited. PZR Pressure Control PZR spray Varied. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 606 Audit Question A-NonLOCA.LTR-46 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event (i.e., system pressures for overheating events, MCHFR for overcooling events). Consequently, sensitivity studies are performed to identify cases with the highest pressure responses for this overheating event. The maximum RCS pressure is limited by the RSV lifting, with its applied lift pressure bias, and therefore many cases may have similar peak pressures. Extensive sensitivity studies are not required to investigate the small differences between those cases. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies performed to identify the highest primary and secondary side pressures, varying primary conditions, break size and location, and steam generator heat transfer and pressure, are presented in Table 7-57. Unless otherwise noted, breaks are located at the bioshield, RCS flows are initialized at the minimum value, and losses of normal AC power occur at event initiation. The results of the sensitivity studies indicate that peak primary pressures occur during a break coincident with a loss of AC power; under these conditions the lowest-setpoint RSV operates to mitigate peak pressurization well below the acceptance criterion of 2310 for this representative NPM. Initial condition biasing contributions are secondary compared to the pressure response to the total loss of heat sink and multiple cases have similar peak pressures. With Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c) Table 7-56 Initial conditions, biases, and conservatisms - feedwater line break (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 607 respect to maximum secondary side SG pressure, maximum SG pressure occurs during a 10 percent split break just outside of containment coincident with a loss of AC power when initial SG pressure is biased to the high condition. The peak secondary side pressures remain well below the design pressure (2100 for this representative NPM). RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-57 Representative sensitivity studies - feedwater line breakNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 608 7.2.13 Uncontrolled Control Rod Assembly Bank Withdrawal from Subcritical or Low Power Startup Conditions The methodology used to simulate a postulated uncontrolled CRA bank withdrawal from subcritical or low power startup conditions for an NPM, and an evaluation of the acceptance criteria listed for an AOO in Table 7-4, are presented below. The range of initial power levels associated with low power startup conditions for an NPM is based on the low setting for the high power signal (below 15 percent RTP for the example in Table 7-3). When core power reaches the low setting level, a hold point is established to alter the high power setting. Thus, low power startup conditions exist until reactor power reaches the low setting level. 7.2.13.1 General Event Description and Methodology The limiting event consequences to an uncontrolled CRA bank withdrawal from subcritical or low power startup conditions typically (for most PWR designs) occur for cases with very low initial power levels (~1 Watt). The primary reason for this behavior is the flux rate signals associated with the source range and intermediate range are typically either not safety related or not of sufficient quantity to adequately address single failures. An NPM, however, incorporates a safety related signal for each of these channels from each core quadrant into the MPS. Consequently, the limiting event consequences for an NPM typically occur for cases with higher initial power levels. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 A spectrum of constant reactivity insertion rates is evaluated, (( }}2(a),(c). These reactivity insertion rates encompass the credible range resulting from a single control bank withdrawal. If necessary, this range is supplemented to include the reactivity insertion rates associated with an inadvertent decrease in boron concentration event (Section 7.2.16). Two event scenarios with different protection schemes are evaluated to determine which scenario produces the limiting event consequences. The first scenario arises when the high count rate signal is available because the intermediate range channel does not have an established signal. In this instance, the high power rate signal is not active (below 15 percent RTP in the Table 7-3 example), so core protection is provided by the high count rate signal and the startup rate (source range) signal. (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 609 The second scenario arises when the high count rate signal is not available because the intermediate range channel has an established signal. In this instance, the high power rate signal is not active (below 15 percent RTP in the Table 7-3 example), thus core protection is provided by the high power signal and the startup rate (intermediate range) signal. The event scenario with the highest core power typically corresponds to the initial power level and reactivity insertion rate that cause the high power signal (low setting) and the startup rate (intermediate range) signal to occur at nearly the same time. This scenario is typically limiting because it represents the maximum rate of power change at the maximum core power. If the initial power is increased, the reactor trips on the high power signal but at a slower rate of power increase. Similarly, if the reactivity insertion rate is increased, the reactor trips on the startup rate but at a lower core power. Before initiating an approach to critical, the reactor coolant is heated to the minimum temperature for criticality. The heating of the reactor coolant is performed by the Module Heatup System (MHS) via the CVCS. At least one feedwater pump is operating at these RCS temperatures. Since feedwater flow is provided to both SGs, it may continue to provide decay heat removal following an uncontrolled CRA bank withdrawal from subcritical or low power startup conditions. If normal feedwater flow is not available, then depending on the point at which the event occurs in the startup process, the flooded containment or the DHRS provides decay heat removal. The peak power and duration of the power spike are not sufficient to cause a significant temperature or pressure increase. Hence, the maximum power and minimum CHFR occur shortly after reactor trip while the RCS pressure and MS pressure do not challenge the relevant acceptance criteria. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as identified in Table 7-60on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR or fuel centerline temperature. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. Table 7-58 lists the relevant acceptance criteria, SAF, and LOP scenarios. Table 7-58 Acceptance criteria, single active failure, loss of power scenarios - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion MCHFR and maximum fuel centerline temperature. MCHFR and maximum fuel centerline temperature are challenged during this reactivity anomaly event.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 611 7.2.13.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-60 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-60 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Nominal. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Nominal. Initial PZR level Nominal. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Biased to BOC conditions. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 612 Reactivity insertion rate VariedSpectrum. RCS Temperature Control Automatic rod control N/A. Boron concentration Not credited. PZR Pressure Control PZR spray Enabled. PZR heaters Enabled. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves N/A. Turbine bypass valves Enabled. Feedwater and Turbine Load Control feedwater pump speed N/A. CNV Pressure Control CNV evacuation system Enabled.

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}}2(a),(c) Table 7-60 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal from subcritical or low power startup conditions (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 613 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generallystudies are performed to identify cases for lowest MCHFR and highest fuel centerline temperature for this reactivity event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity study results typicallyFor example, representative sensitivity studies for the initial core power and reactivity insertion rate are presented in Table 7-61. The results demonstrate the lack of challenging MCHFR values predicted for this event by NRELAP5, which is further supported by the values predicted with the approved subchannel methodology. (( }}2(a),(c) RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-61 Representative sensitivity studies - uncontrolled control rod bank withdrawal from subcritical or low power startup conditionsNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 614 7.2.14 Uncontrolled Control Rod Assembly Bank Withdrawal at Power The methodology used to simulate a postulated uncontrolled control rod assembly (CRA) bank withdrawal at power for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.14.1 General Event Description and Methodology As stated in Section 7.2.13, low power startup conditions exist for an NPM until reactor power reaches the low setting level (15 percent RTP for the example in Table 7-3). Accordingly, the uncontrolled CRA bank withdrawal at power event extends from the low setting level to HFP. The withdrawal of the control bank causes a reactivity insertion that increases reactor power and leads to a rise in coolant temperature, pressurizer level, and RCS pressure. Reactivity feedback from the rising fuel temperature partially counteracts the reactivity insertion, slowing the power increase, which continues until the system trips on high power, high power rate, high pressurizer pressure, or high RCS temperatures. The maximum power and minimum MCHFR occur just after the resulting scram, while the peak primary pressure occurs some time later, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established by the end of the transient. Following reactor trip and subsequent turbine trip, the turbine bypass to the condenser opens to control the RCS temperature. However, the actions of the turbine bypass system are not credited, so as to minimize heat removal by the secondary side. Although turbine load is an input to the feedwater controller, no changes are made to this controller because the RCS responses are not sufficient to affect feedwater control. Table 7-61 Representative sensitivity studies - uncontrolled control rod bank withdrawal from subcritical or low power startup conditionsNot Used (Continued) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 615 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The limiting MCHFR typically occurs for a reactivity insertion rate that results in reactor trip on core power, pressurizer pressure, and RCS riser temperature signals at approximately the same time. These conditions typically arise for events initiated with lower reactivity insertion rates because higher reactivity insertion rates cause the MPS to trip much earlier on high power rate. The earlier reactor trip reduces the energy added to the reactor coolant, thereby producing a higher MCHFR. The range of reactivity insertion rates considered is sufficient to identify the point of transition (( }}2(a),(c) Sensitivity studies are performed on a variety of parameters, as identified in Table 7-64necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. The maximum fuel centerline temperature typically occurs when core power exceeds its analytical limit. This condition typically arises for events initiated from full power with the highest reactivity insertion rate as determined from the resulting bank worth and control rod step speed. A spectrum of constant reactivity insertion rates is evaluated. These reactivity insertion rates encompass the credible range resulting from a single control bank withdrawal. If necessary, this range is supplemented to cover the reactivity insertion rates associated with an inadvertent decrease in boron concentration event (Section 7.2.16). Table 7-62 lists the relevant acceptance criteria, SAF, and LOP scenarios. 7.2.14.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-63. Table 7-62 Acceptance criteria, single active failure, loss of power scenarios - uncontrolled control rod bank withdrawal at power Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion MCHFR and maximum fuel centerline temperature MCHFR and maximum fuel centerline temperature are challenged during this reactivity anomaly event. No single failure The challenging cases typically occur when all equipment is operational. No loss of power The challenging cases typically occur when AC power is available for the event duration.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 616 7.2.14.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-64 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature. Audit Question A-NonLOCA.LTR-59 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-63 Acceptance criteria - uncontrolled control rod bank withdrawal at power Acceptance Criteria Discussion Primary pressure This criterion is not an acceptance criterion listed in Section 15.4.2 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for primary pressure compared to other AOOs. Secondary pressure This criterion is not an acceptance criterion listed in Section 15.4.2 of the SRP (Reference 15). The analysis shows an NPM does not introduce more challenging conditions for secondary pressure compared to other AOOs. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power Parameter Bias / Conservatism Basis (( Initial reactor power Varied. RTP biased upwards to account for measurement uncertainty. }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 617 Initial RCS average temperature Varied. Biased to the high condition. Initial RCS flow rate Varied. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the low condition. MTC Biased to BOC conditions. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Reactivity insertion rate Spectrum.Varied Maximum Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 619 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generallystudies are performed to identify cases for lowest MCHFR for this reactivity event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies for initial core power, reactivity insertion rate, pressurizer pressure, pressurizer level, reactor coolant average temperature, pressurizer spray flow, pressurizer heater status, letdown status, and loss of power are presented in Table 7-65. The results demonstrate the two most limiting MCHFRs occur at initial power levels of 75 percent RTP and above, with AC power available for the event duration, the RCS average temperature biased low, and pressurizer spray modeled to delay reactor trip on high pressurizer pressure. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c) Table 7-65 Representative sensitivity studies - uncontrolled control rod bank withdrawal at powerNot Used (( }}2(a),(c) Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 620 7.2.15 Control Rod Misoperation The methodology used to simulate a postulated control rod misoperation for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.15.1 General Event Description and Methodology The rod control system is used to move (insert or withdraw) the control rod assemblies (CRAs) in response to an operator action or an automatic control. Since these transients are initiated by a malfunction in the rod control system, a variety of reactivity related conditions can result. Specific reactivity conditions for an NPM include: 1) withdrawing a single CRA; 2) dropping one or more CRAs; or,

3) leaving one or more CRAs behind when inserting or withdrawing a control bank. The consequences for each of these reactivity conditions are discussed below.

Table 7-66 lists the relevant acceptance criteria, single active failure, and loss of power scenarios. Table 7-65 Representative sensitivity studies - uncontrolled control rod bank withdrawal at powerNot Used (Continued) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 621 Withdrawal of a Single CRA The withdrawal of a single CRA causes a reactivity insertion that increases reactor power and leads to a rise in coolant temperature, pressurizer level, and RCS pressure. Feedback from the rising fuel temperature is not sufficient to counteract the reactivity insertion, so the power increases until the system trips on high power, high power rate, high pressurizer pressure, or high RCS temperatures. The maximum power and minimum MCHFR occur just after the resulting scram, while the peak primary pressure occurs some time later, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established at the end of the transient. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The limiting MCHFR typically occurs for a reactivity insertion rate that results in reactor trip on core power, pressurizer pressure, and RCS temperature signals at approximately the same time. These conditions arise for events initiated from partial power with lower reactivity insertion rates because higher reactivity insertion rates cause the MPS to trip much earlier on high power rate. The earlier reactor trip reduces the energy added to the reactor coolant, thereby producing a higher MCHFR. The asymmetry associated with the core power response causes the ex-core detectors to respond differently for each quadrant. Consequently, the range of reactivity insertion rates considered is sufficient to identify the point of transition (( }}2(a),(c)to the high power rate signal (with power related trips using the lowest reading ex-core detector based on the minimum after to before event initiation ratio of the radial peaking factors for the outer row of fuel assemblies). Sensitivity studies are performed on a variety of parameters, as necessaryindicated in Table 7-68, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. In contrast to the uncontrolled CRA bank withdrawal at power event (Section 7.2.14), the highest reactivity insertion rate as determined from the resulting rod worth and control rod step speed is significantly lower for the withdrawal of a single CRA event. Audit Question A-NonLOCA.LTR-5 Note that instead of using case-specific values for rod worth and parameters associated with the asymmetry, more conservative bounding input values can be used. The use of bounding input values provides a conservative analysis simplification for the withdrawal of a single CRA event. In addition, the bounding input values can then be used to compare this event to the dropping of one or more CRAs event described below. Dropping One or More CRAs Based on the minimum worth at any time during the cycle for a given core power, i.e., with the control bank positioned at the PDIL, dropping a single CRA causes a reactivity insertion that decreases reactor power. Feedback from the decreasing fuel temperature and the actions of the rod control system to restore power are

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 622 generally not sufficient to counteract the reactivity insertion, so the power decreases until the system trips on high power rate. For event scenarios without a return to power, the maximum core power, peak primary pressure, and MCHFR occur at event initiation. The peak secondary system pressure occurs some time after the scram, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established at the end of the transient. The potential for a return to power exists only for events initiated from less than RTP because the reduced worth of the dropped rod gives the rod control system time to act. The corresponding MCHFR for a dropped rod event with a return to power is typically greater than the MCHFR for events initiated from HFP. Hence, the limiting MCHFR cases typically occur at HFP conditions. Following reactor trip and subsequent turbine trip, the turbine bypass to the condenser opens to control the RCS temperature. However, the actions of the turbine bypass system are not credited, so as to minimize heat removal by the secondary side. Although turbine load is an input to the feedwater controller, no changes are made to this controller because the RCS responses are not sufficient to affect feedwater control. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 The asymmetry associated with the core power response causes the ex-core detectors to respond differently for each quadrant. The power input to the high power rate signal uses the highest reading ex-core detector, multiplying the core average power by the maximum after drop to before drop ratio of the radial peaking factors for the outer row of fuel assemblies. Sensitivity studies are performed, as identified in Table 7-70 on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. The maximum fuel centerline temperature typically occurs at event initiation for those event scenarios with an immediate reactor trip. If the event scenario has a return to power, the maximum fuel centerline temperature is typically bounded by the fuel centerline temperature at HFP because the associated power peak is less than full power. Audit Question A-NonLOCA.LTR-5 As an alternative to performing a system transient analysis, the MCHFR and linear heat generation rate of the dropped rod event can be confirmed to be bounded by other events. Most rod drops result in a reactor trip on high power rate because of the immediate decrease in power from the dropped rod. Figure 7-3 shows the decrease in power from the dropped rod (( }}2(a),(c) as a function of rod worth for a variety of initial power levels. The figure is based on a representative NPM core design and also overlays a representative high power rate trip. (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 628 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-68 Initial conditions, biases, and conservatisms - control rod misoperation, single control rod assembly withdrawal Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Varied. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied. Initial PZR level Varied. Initial feedwater temperature Nominal. Initial fuel temperature Biased to the low condition. MTC Biased to BOC conditions. Kinetics Biased to BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Reactivity insertion rate Spectrum.Varied. RCS Temperature Control Automatic rod control N/A. }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 629 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generallystudies are performed to identify cases for lowest MCHFR for this reactivity event. Boron concentration Not credited. PZR Pressure Control PZR spray Varied. PZR heaters Varied. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c) Table 7-68 Initial conditions, biases, and conservatisms - control rod misoperation, single control rod assembly withdrawal (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 630 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies for initial core power, reactivity insertion rate, pressurizer pressure, pressurizer level, reactor coolant average temperature, pressurizer spray flow, pressurizer heater status, letdown status, and loss of power are presented in Table 7-69 for the withdrawal of a single CRA event. The results demonstrate the MCHFR occurs at an initial power level of 75 percent RTP, with AC power available for the event duration, the RCS average temperature biased low, the pressurizer pressure biased low, the pressurizer level biased low, and spray modeled to delay reactor trip on high pressurizer pressure. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-69 Representative sensitivity studies - control rod misoperation, single control rod assembly withdrawalNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 631 Dropping One or More CRAs The biases and conservatisms presented in Table 7-70 are considered in order to identify the bounding transient simulation for MCHFR and maximum fuel centerline temperature for the dropped CRA(s) event. If the alternative bounding approach is used, where cases that result in early trip are bounded by steady-state condtions and cases that do not result in early trip are bounded by the single rod withdrawal, the biases and conservatisms in Table 7-70 associated with system transient analysis are not applicable. However, the alternative bounding approach considers combinations of initial core power, dropped CRA worth, core time-in-life, axial offset, flow, and temperature to screen the cases into groups and compare the non-tripped cases to single rod withdrawal analysis limits. Table 7-69 Representative sensitivity studies - control rod misoperation, single control rod assembly withdrawalNot Used (Continued) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 633 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generallystudies are performed to identify cases for lowest MCHFR for this reactivity event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies for the initial core power, dropped CRA worth, and core time-in-life are presented in Table 7-71 for the dropped CRA(s) event. The results demonstrate the MCHFR occurs at full power conditions. Boron concentration Not credited. PZR Pressure Control PZR spray Nominal. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled. Table 7-70 Initial conditions, biases, and conservatisms - control rod misoperation, dropped control rod assemblies (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 634 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.16 Inadvertent Decrease in Boron Concentration The methodology used to simulate an inadvertent decrease in boron concentration for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.16.1 General Event Description and Methodology The boric acid blend system incorporated into the NuScale plant design permits the operator to control the boron concentration of the reactor coolant via the charging fluid chemistry. While the NuScale plant design incorporates both automatic and manual controls, strict administrative procedures govern the process for adjusting the boron concentration of the reactor coolant. These administrative procedures establish limits on the rate and duration of the dilution. The primary means of causing an inadvertent decrease in boron concentration is failure of the blend system, either by controller or mechanical failure, or operator error. The event is terminated by isolating the source for the diluted water, i.e., by closing the demineralized water system (DWS) isolation valves. For Mode 1 plant operating conditions, the perfect mixing model and the wave front model are both evaluated. The perfect mixing model is evaluated for Mode 1 operating conditions because it provides a slower reactivity insertion rate, delaying detection, potentially allowing further loss of shutdown margin. The wave front model is physically conservative because it assumes the maximum amount of reactivity as the diluted slug of water sweeps through the core. This model does not assume any axial blending to ensure that this reactivity insertion rate is maximized. For all other operating modes where boron dilution is allowed and limited mixing exists, a wave front model is used. These mixing models are Table 7-71 Representative sensitivity studies - control rod misoperation, dropped control rod assembliesNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 642 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 (( }}2(a),(c) RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-75 Representative results - inadvertent decrease in boron concentration in Mode 1 at hot full powerNot Used Mode 1 (HFP) 1 CVCS Pump 2 CVCS Pumps Dilution Flow Rate (gpm) 25 50 Initial boron concentration (ppm) 1400 1400 Final boron concentration (ppm) 1196 1196 Perfect mixing reactivity insertion rate (pcm/s) 0.50 1.00 Reactor trip / DWS isolation time (s) 161 78 DWS isolation valves closed (s) 166 83 Shutdown margin at time of isolation (pcm) Greater than 1944 Greater than 1912 Shutdown margin lost assuming no isolation (s) 4092 (68.2 min) 2046 (34.1 min) Table 7-76 Representative results - inadvertent decrease in boron concentration in Mode 1 at 25 percent rated thermal powerNot Used Mode 1 (25 percent RTP) 1 CVCS Pump Dilution flow rate (gpm) 25 Initial boron concentration (ppm) 1800 Final boron concentration (ppm) 1596 Perfect mixing reactivity insertion rate (pcm/s) 0.65 Wave front reactivity insertion rate (pcm/s) 19.1 Wave front initial reactivity step (pcm) 91 Shutdown margin lost assuming no isolation (s) 3120 (52 min) Table 7-77 Representative results - inadvertent decrease in boron concentration in Mode 1 at hot zero powerNot Used Mode 1 (HZP, 1 MWt) 1 CVCS Pump Dilution Flow Rate (gpm) 25 Initial boron concentration (ppm) 1800 Final boron concentration (ppm) 1596 Initial reactivity step (pcm) 685 Wave front reactivity insertion rate (pcm/s) 17.3 Duration of reactivity insertion rate for each wave (s) 39.5 Initial wave front reaches core inlet (s) 900 Initial wave front reaches core exit (s) 939.5 Reactor trip / DWS isolation time(1) (s) (wave front at core exit + UCRWS trip time) Approximately 2028 DWS isolation valves closed (s) Approximately 2033

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 643 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.17 Chemical and Volume Control System Malfunction that Increases Reactor Coolant System Inventory The methodology used to simulate a postulated CVCS malfunction that increases RCS inventory for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. Shutdown margin at time of isolation (pcm) Greater than 697 Shutdown margin lost assuming no isolation (s) (critical wave at core inlet) 4200 (70 min)

1. Conservatively estimated to occur upon arrival of the subsequent wave Table 7-78 Representative results - inadvertent decrease in boron concentration in Mode 2Not Used Mode 2 (Coolant Flow 1.7 ft3/s) 1 CVCS Pump Dilution Flow Rate (gpm) 25 Initial boron concentration (ppm) 785.5 Final boron concentration (ppm) 600 Initial reactivity step (pcm) 329 Wave front reactivity insertion rate (pcm/s) 8.32 Duration of reactivity insertion rate for each wave (s) 39.5 Reactor trip/DWS isolation wave at core inlet (s) 5280 (88 min)

DWS isolation valves closed (s) 5475 Shutdown margin at time of isolation (pcm) Greater than 517 Shutdown margin lost assuming no isolation (s) (critical wave at core inlet) 7440 (124 min) Table 7-79 Representative results - inadvertent decrease in boron concentration in Mode 3Not Used Mode 3 (Coolant Flow = 1.7 ft3/s) 1 CVCS Pump Dilution flow rate (gpm) 25 Initial boron concentration (ppm) 813 Final boron concentration (ppm) 650 Initial reactivity step (pcm) 378 Wave front reactivity insertion rate (pcm/s) 9.55 Duration of reactivity insertion rate for each wave (s) 39.6 Reactor trip / DWS isolation wave at core inlet (s) 4200 (70 min) DWS isolation valves closed (s) 4395 Shutdown margin at time of isolation (pcm) Greater than 613 Shutdown margin lost assuming no isolation (s) (critical wave at core inlet) 6360 (106 min) Table 7-77 Representative results - inadvertent decrease in boron concentration in Mode 1 at hot zero powerNot Used (Continued) Mode 1 (HZP, 1 MWt) 1 CVCS Pump

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 644 7.2.17.1 General Event Description The NuScale reactor does not have a pumped ECCS, therefore, the unplanned increase in RCS inventory event is caused by a malfunction of the CVCS makeup pumps or pressurizer level control system. If borated water at the same concentration of the primary system is added to the RCS, the addition of large amounts of water to the primary system typically generates a reactor trip on high pressurizer (PZR) water level or high PZR pressure. Table 7-80 lists the relevant acceptance criteria, SAF, and LOP scenarios. The malfunction is assumed to isolate letdown and actuate both makeup pumps (each flowing at their maximum capacity), causing an unplanned increase in RCS inventory. The limiting pressure response occurs when the event is initiated from full power conditions, and the initial conditions are biased in the conservative directions. The increase in RCS inventory event is typically terminated by CVCS isolation on high PZR level. (The CVCS containment isolation valves are dual safety related valves.) RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies to identify the challenging conditions are performed, as necessaryidentified in Table 7-82, to identify the case(s) with the potentially limiting system pressures. Cases that include the effects of CVCS recirculation, which increases the flow into the RCS, are also analyzed. 7.2.17.2 Acceptance Criteria Evaluation of the most challenging case(s) relative to the acceptance criteria is presented in Table 7-81. Table 7-80 Acceptance criteria, single active failure, loss of power scenarios - reactor coolant system inventory increase Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion Primary pressure, secondary pressure System pressures are challenged during this mass addition event. No single failure The CVCS is isolated via dual safety-related isolation valves. If one of the isolation valves were to fail, the other CVCS isolation valve would provide system isolation. No loss of power Continued operation of the CVCS typically maximizes system pressures.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 647 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event, including cases with the highest pressure responses for this inventory increase event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, sensitivity studies performed to identify the highest primary and secondary side pressures, varying primary conditions, makeup temperature, and spray availability are presented in Table 7-83. The results indicate that peak primary pressure occurs when initial primary temperature and pressure are biased to the low condition, and makeup temperature is biased to the high condition as the RTS actuates on high pressurizer pressure and the lowest-setpoint RSV operates to mitigate peak pressurization well below the acceptance criterion of 2310 for this representative NPM. With respect to maximum secondary side SG pressure, maximum SG pressure occurs when spray is used to delay the RTS actuation until high pressurizer level is reached. Turbine bypass valves Disabled. Feedwater and Turbine Load Control Feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

1. ((

}}2(a),(c)

2. These inputs, in conjunction with least negative Doppler temperature coefficient, are selected to maximize the power response (if any) induced by the addition of colder CVCS water. However, since this event is driven by mass addition, reactivity effects prior to RTS actuation (if any) are small when compared to the pressurization associated with the increase in primary inventory.

Table 7-82 Initial conditions, biases, and conservatisms - reactor coolant system inventory increase (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 648 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.18 Failure of Small Lines Outside Containment The methodology used to simulate a postulated failure of a small line connected to the primary coolant system outside of containment for an NPM, and an evaluation of the acceptance criteria for an infrequent event listed in Table 7-4, are presented below. A postulated break in a small line carrying primary coolant is typically only evaluated for radiological consequences. Neither the plant design nor the use of natural circulation flow for an NPM introduces a more challenging condition for other acceptance criteria. Evaluation of postulated small line breaks within the context of a Table 7-83 Representative sensitivity studies - reactor coolant system inventory increaseNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 655 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies for break location and size to determine impact on mass release and iodine spiking time are presented in Table 7-87 and Table 7-88, respectively. The Table 7-87 results indicate the mass release is maximized with a break in the letdown line of 100 percent area, AC power lost at event initiation, RCS average temperature biased high, and a break in the makeup line of 100 percent area coincident with reactor trip. The Table 7-88 results indicate the iodine spiking duration (elapsed time from reactor trip to CVCS isolation) is maximized with a break in the makeup line, AC power lost at event initiation, RCS average temperature biased high, and no break in the letdown line coincident with reactor trip. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.

1. ((

2. }}2(a),(c) Table 7-86 Initial conditions, biases, and conservatisms - breaks in small lines carrying primary coolant outside containment (Continued) Parameter Bias / Conservatism Basis(1) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 656 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-87 Representative break, time in life, power, flow, and temperature sensitivity study for mass release - breaks in small lines carrying primary coolant outside containmentNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 657 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 7.2.19 Steam Generator Tube Failure The methodology used to simulate a postulated failure of a steam generator tube for an NPM, and an evaluation of the acceptance criteria for an accident listed in Table 7-4, are presented below. A postulated failure of a steam generator tube is typically only evaluated for radiological consequences. Neither the steam generator design nor the use of natural circulation flow for an NPM introduce a more challenging condition for other acceptance criteria. Table 7-88 Representative break, time in life, power, flow, and temperature sensitivity study for iodine spiking time - breaks in small lines carrying primary coolant outside containmentNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 663 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. PZR Pressure Control PZR spray Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled. (( 1.

2.

}}2(a),(c) Table 7-91 Initial conditions, biases, and conservatisms - steam generator tube failure (Continued) Parameter Bias / Conservatism Basis (1) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 664 RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 For example, representative sensitivity studies for break characteristics, core time-in-life, SG pressure, SG tube plugging, loss of power, single failure, feedwater temperature and reactor coolant temperature for a steam generator tube failure with respect to mass release and iodine spiking duration are presented in Table 7-92. The results indicate the mass release and spiking time are maximized with a double-ended guillotine break at the top of the SG tube, AC power available for the event duration, and a single active failure of the primary MSIV on the affected SG to close. RAI 10297 Question NonLOCA.LTR-31, 32, 46, 56, 65 Table 7-92 Representative break characteristics, initial conditions, loss of power, and single active failure sensitivity study - steam generator tube failureNot Used (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 665 Table 7-92 Representative break characteristics, initial conditions, loss of power, and single active failure sensitivity study - steam generator tube failureNot Used (Continued) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 537 7.2.1.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-7 are considered in identifying a bounding transient simulation for MCHFR.53 Audit Question A-NonLOCA.LTR-28 RAI Questions A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition Initial PZR level Varied.Biased to the high condition. Initial feedwater temperature Varied. Initial fuel temperature Nominal. Moderator Temperature Coefficient (MTC) Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Varied. SG heat transfer Nominal RSV lift setpoint Nominal }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 547 Audit Questions A-NonLOCA.LTR-28, A-NonLOCA.LTR-65 RAI Questions, A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level VariedBiased to the high condition. Initial feedwater temperature Varied. Initial fuel temperature Nominal. MTC Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the low condition. }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 558 Audit Question A-NonLOCA.LTR-28 RAI Questions A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level Varied.Biased to the high condition. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Biased to the high condition. SG heat transfer Nominal RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Steam flow increase Varied }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 559 RCS Temperature Control Automatic rod control Varied. Boron concentration Not credited. PZR Pressure Control PZR spray Varied.Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves N/A. Turbine bypass valves N/A. Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 566 RCS Temperature Control Automatic rod control Enabled. (MCHFR) Boron concentration Not credited. PZR Pressure Control PZR spray Varied.Disabled. PZR heaters EnabledNominal. PZR Level Control Charging Not credited. Letdown Disabled. Table 7-24 Initial conditions, biases, and conservatisms - steam line break (Continued) Parameter Bias / Conservatism Basis(1) (( }}2(a),(c)

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 567 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, sensitivity studies are performed to identify cases with the lowest CHFR response and challenging mass releases for this overcooling event. Steam Pressure Control Turbine throttle valves EnabledVaried. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. CNV Pressure Control CNV evacuation system Enabled. (( }}2(a),(c) Table 7-24 Initial conditions, biases, and conservatisms - steam line break (Continued) Parameter Bias / Conservatism Basis(1) (( }}2(a),(c)

RAIO-177581 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Affidavit of Mark W. Shaver, AF-177583

AF-177583 Page 1 of 2

NuScale Power, LLC AFFIDAVIT of Mark W. Shaver I, Mark W. Shaver, state as follows: (1) I am the Director of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale. (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying Request for Additional Information response reveals distinguishing aspects about the response by which NuScale develops its NuScale Power, LLC Response to NRC Request for Additional Information (RAI No. 10297 R1, Question NonLOCA.LTR-31, 32, 46, 56, 65) on the NuScale Standard Design Approval Application. NuScale has performed significant research and evaluation to develop a basis for this response and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScales competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScales intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed response to NRC Request for Additional Information RAI 10297 R1, Question NonLOCA.LTR-31, 32, 46, 56, 65. The enclosure contains the designation Proprietary at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, (( }} in the document.

AF-177583 Page 2 of 2 (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScales technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on January 09, 2025. Mark W. Shaver}}