ML24353A333

Nuscale Power, LLC - Response to NRC Request for Additional Information No. 037 (RAI-10357 Ri) on the Newscale Standard Design Approval Application

ML24353A333
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Site:	05200050
Issue date:	12/18/2024
From:	Shaver M NuScale
To:	Office of Nuclear Reactor Regulation, Document Control Desk
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RAIO-177177
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RAIO-177177 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com December 18, 2024 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. 037 (RAI-10357 R1) on the NuScale Standard Design Approval Application

REFERENCE:

NRC Letter to NuScale, Request for Additional Information No. 037 (RAI-10357 R1), dated October 17, 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-10357 R1:

15.1.1-7 is the proprietary version of the NuScale Response to NRC RAI No. 037 (RAI-10357 R1, Question 15.1.1-7). 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 December 18, 2024.

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

RAIO-177177 Page 2 of 2 12/18/2024 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 Stacy Joseph, Senior Project Manager, NRC

NuScale Response to NRC Request for Additional Information RAI-10357 R1, Question 15.1.1-7, Proprietary Version : NuScale Response to NRC Request for Additional Information RAI-10357 R1, Question 15.1.1-7, Nonproprietary Version : Affidavit of Mark W. Shaver, AF-177178

RAIO-177177 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-10357 R1, Question 15.1.1-7, Proprietary Version

RAIO-177177 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-10357 R1, Question 15.1.1-7, Nonproprietary Version

Response to Request for Additional Information Docket: 052000050 RAI No.: 10357 Date of RAI Issue: 10/17/2024 NRC Question No.: 15.1.1-7 Issue As described in FSAR Section 15.1, malfunctions resulting in an increase in primary heat removal by the secondary system cause the moderator density to increase. Increased moderator density in the downcomer region affects the excore detector signal. NuScale models an increased high power level trip setpoint during overcooling events to account for this factor.

NRC staff audited (ML23067A300) calculations supporting this modification to the high power level trip setpoint and found that the modification applied to the safety analysis is larger than the calculated value.

Changes in downcomer temperature will also affect reactor trip on the high power rate signal.

However, the FSAR and methodology documentation, including information provided during the audit, does not contain sufficient information concerning modeling of the high power rate trip setpoint during overcooling events.

During the audit, NuScale clarified that the high power rate trip is assumed to operate in the sensitivity studies on the decrease in feedwater temperature event discussed in FSAR Section 15.1.1 and the increase in steam flow event discussed in FSAR Section 15.1.3. It is unclear whether inferences made from sensitivity studies in which the high power rate signal results in a reactor trip are applicable to limiting cases that do not model operation of this trip, since terminating the transient at an earlier time may obscure trends resulting from variation of other parameters (e.g., initial power level) in these sensitivity studies.

Additionally, the staff noted during its audit that these sensitivity studies do not account for the effect of changes in downcomer moderator temperature and density on the high power rate trip.

It is unclear to NRC staff whether this approach will result in an accurate or conservative assessment of reactor trip time.

Information Requested Revise calculations supporting the decrease in feedwater temperature and increase in steam flow events such that modeled reactor trip signals are consistent between the limiting case and NuScale Nonproprietary NuScale Nonproprietary

sensitivity studies, unless the sensitivity study is performed to examine modeling of a trip signal.

If the high power rate trip will be credited in limiting cases, confirm that the modeled change in excore detector signal from a change in downcomer temperature is conservative or accurate by providing, for limiting cases, the magnitude of the modification to the excore detector signal or trip setpoint, trip time, and by showing progressions of downcomer temperature, core average temperature, actual reactor power, and indicated reactor power, and other parameters as necessary such that the phenomena leading to the high power rate trip are clear.

Additionally, update portions of the licensing basis discussing the methodology for evaluation of overcooling transients (i.e., the non-LOCA LTR or a subsection of FSAR 15.0) to specify how the evaluation model accounts for changes in excore detector signal caused by changes in downcomer temperature when assessing high power rate trips. Provide justification that this methodology results in a conservative or accurate trip time. If the magnitude of this effect is biased in a specific direction, provide justification that trip time will be accurate or conservative in circumstances in which the effect encourages a high power rate trip (e.g., downcomer temperature is increasing when a positive rate trip is assessed) and discourages a high power rate trip (e.g., downcomer temperature is decreasing when a positive rate trip is assessed), as information provided during the audit does not address all of these circumstances. If justification for this biasing is based on a specific event progression or timing of the high power rate trip, state this basis such that conservatism or appropriateness of the approach for overcooling events is clear. Specify any modeling details (e.g., method of modeling this effect such as a scalar multiplication on core power, formulas used to assess the high power rate trip) needed to support justification that the specific method is conservative, as information provided during the audit does not sufficiently specify this information.

NuScale Response:

Executive Summary NuScale has revised the analyses of the decrease in feedwater temperature and increase in steam flow to reasonably and conservatively address the impact of decreasing downcomer temperature on power-related reactor trip signals. The revised analyses are provided in the Chapter 15 electronic reading room (eRR) for NRC audit. The response describes how the impact of decreasing downcomer temperature on power-related reactor trip signals is treated and why the treatment is conservative. Sensitivity results demonstrating the reasonableness and conservatism of the approach are also provided. The Final Safety Analysis Report (FSAR)

Table 15.0-7 is revised as indicated in the attached markups to provide details of the approach.

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Previously-provided markups to the topical report TR-0516-49416-P, Revision 4, Non-Loss-of-Coolant Accident Analysis Methodology, are supplemented with additional markups to be consistent with the approach.

Revised Chapter 15 Calculations EC-104467, Revision 0, NPM-20 Decrease in Feedwater Temperature Transient Analysis (previously provided in the Chapter 15 eRR for NRC audit), ((2(a),(c) EC-104467, Revision 0, ((

}}2(a),(c) EC-104467, Revision 1, has been issued to address the NRC question, is provided in the eRR, and is discussed further in this response.

EC-120241, Revision 0, NPM-20 Increase in Steam Flow/Inadvertent Opening of Steam Generator Relief or Safety Valve Analysis (previously provided in the Chapter 15 eRR for NRC audit), ((

}}2(a),(c) EC-120241, Revision 0, (( 
}}2(a),(c) EC-120241, Revision 1, has been issued to address the NRC question, is provided in the eRR, and is discussed further in this response.

EC-104467, Revision 1, and EC-120241, Revision 1, ((

}}2(a),(c) reactor trip signals are modeled to consider the decalibration effect caused by decreasing downcomer temperature. The modeling of the power-related reactor trips is described in the following paragraphs.

Treatment of High Power Reactor Trip The high power reactor trip analytical limit is 115 percent of rated thermal power (RTP) per FSAR Table 15.0-7. There is a footnote to this entry that is modified as indicated in the markup provided with the response to audit question A-15.1.1-3 to state The overcooling event analyses account for decreased power detection due to decreasing downcomer temperature. The reduction is 7% for downcomer temperature decreases up to 10°F and is scaled upwards from 7% by 0.7%/°F for downcomer temperature decreases beyond 10°F. The response to audit question A-15.1.1-3 identifies that the 7 percent RTP for the first 10 degrees F and the 0.7 NuScale Nonproprietary NuScale Nonproprietary

percent RTP per degree F beyond 10 degrees F (( }}2(a),(c) With this treatment, the response to audit question A-15.1.1-3 identifies that the high power reactor trip (( }}2(a),(c) In EC-104467, Revision 1, and EC-120241, Revision 1, the modeling of the decalibration effect still results in a minimum 7 percent RTP reduction in detected power for the first 10 degrees F. However, beyond 10 degrees F, a factor of approximately 0.6 percent per degree F (( }}2(a),(c) (i.e., rather than 0.7 percent RTP per degree F (( }}2(a),(c) ). Equation 1 below shows this approach. (( }}2(a),(c) Reactor trip occurs when the detected power in Equation 1 exceeds the FSAR Table 15.0-7 analytical limit. The factor of approximately 0.6 percent core power per degree F (( }}2(a),(c) In addition, this treatment results in the same level of conservatism for a downcomer temperature decrease up to 10 degrees F (( }}2(a),(c) and results in a slightly more conservative treatment as downcomer temperature decreases beyond 10 degrees F. For example, using NuScale Nonproprietary NuScale Nonproprietary

Equation 1, (( }}2(a),(c) Because this modeling of the high power reactor trip decalibration effect in EC-104467, Revision 1, and EC-120241, Revision 1, is the same as or more conservative than that previously described in the response to audit question A-15.1.1-3 that was accepted by the NRC for resolution of that audit question, no further justification of this treatment is provided in this response. Treatment of High Power Rate Reactor Trip The high power rate reactor trip analytical limit is +/-7.5 percent per 30 seconds in FSAR Table 15.0-7. There is no current footnote to this entry to describe consideration of the decalibration effect. In EC-104467, Revision 1, and EC-120241, Revision 1, decalibration is considered for the high power rate reactor trip. A factor of approximately 0.6 percent per degree F (( }}2(a),(c) (i.e., the same as discussed above) is applied to determine the decalibration amount. No minimum decalibration amount is applied (i.e., the 7 percent RTP reduction in detected power for the first 10 degrees F used for the high power reactor trip). This treatment is applied to both the instantaneous power and the 30-second average power for determination of when the two powers differ by more than 7.5 percent. Because this modeling of the high power rate reactor trip decalibration effect in EC-104467, Revision 1, and EC-120241, Revision 1, is new and was not previously described in the FSAR (or FSAR markups associated with related audit questions), further justification of this treatment is provided in this response. The use of a multiplier for accounting for decalibration in the high power rate reactor trip was previously proposed in the response to audit question A-15.1.1-7, with 0.7 percent RTP per degree F provided as an example. Justification for why such an approach would be appropriately conservative was previously provided and a summary of that justification is provided here:

The applied decalibration factor of approximately 0.6 percent per degree F (( }}2(a),(c) is not overly conservative. The NRC feedback on the response to audit question A-15.1.1-7 expressed a concern that using a large decalibration factor is not conservative if it causes a negative high power rate reactor trip to occur. NuScale agrees that use of a decalibration factor that is too large could cause a negative high power rate trip to occur based on the cooldown effect of the event alone. However, NuScale is not using an overly conservative NuScale Nonproprietary NuScale Nonproprietary

decalibration factor. (( }}2(a),(c)

The power-averaging of the reactor trip allows time for power to increase in the core. During cooldown events, the decrease in downcomer temperature occurs first. The increased neutron attenuation through the colder (i.e., more dense) downcomer water will cause measured power at the detectors to decrease. It could be postulated that the measured power used for the high power rate reactor trip is therefore decoupled from the core. However, due to the 30-second-averaging of the power, there is time for the colder water from the downcomer to reach the core and cause core power to increase because of moderator feedback. The increase in core power will result in an increase in measured power at the detectors. The 30-second-averaged measured power will not be uncoupled from the core power. Note that although loop transit time is larger for low powers, limiting overcooling events occur from full power where loop transit time is minimized.

Overcooling events consider automatic rod withdrawal. When temperature decreases in the downcomer, the thermocouples in the downcomer will register this change and average temperature (Tavg) will decrease. With automatic rod withdrawal modeled, rods will begin withdrawing due to the decrease in Tavg. Rod withdrawal will result in a power increase in the core, and a resulting increase in measured power at the detectors. As stated before, the 30-second-averaged measured power will not be uncoupled from the core power.

Negative high power rate reactor trips can be ignored. Ignoring the reactor trip actuation from occurrence of a negative high power rate reactor trip, if it is calculated to occur, eliminates the potential that a high power rate reactor trip occurs in a non-conservative manner.

Aside from the concern of negative high power rate reactor trips, addressed above, there is no concern with using a decalibration factor (( }}2(a),(c) for positive high power rate reactor trips. A larger decalibration factor will force a correspondingly larger increase in actual core power before the high power rate reactor trip occurs based on detected power. Delaying trip and allowing larger power levels to be reached is conservative. Therefore, use of a larger multiplier is conservative because it delays a reactor trip relative to using a smaller (or no) decalibration factor.

Use of a scalar decalibration factor most closely approximates the actual physical behavior. As described above, the power detected at the detectors is affected by the neutron attenuation through the downcomer, core power change due to rod movement, NuScale Nonproprietary NuScale Nonproprietary

and core power response to moderator temperature changes. Application of a scalar decalibration factor results in dynamic changes in measured power that are consistent with the expected behavior. The inherent time-averaging of the high power rate reactor trip then ensures that only significant power increases above the time-average will result in a reactor trip in the analysis. Sensitivity Study Supporting Treatment of High Power Rate Reactor Trip To illustrate and supplement the above justifications, a sensitivity study was performed ((

}}2(a),(c) The sensitivity study results (( 
}}2(a),(c) are provided in this response as Table 1.

((

}}2(a),(c)

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Table 1: High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Comparisons of the cases for various parameters are shown in Figure 1 through Figure 5 for the full transient. (( }}2(a),(c) Comparisons of the cases for various parameters focused on the behavior leading up to reactor trip are shown in Figure 6 through Figure 12. The decrease in reactor coolant system (RCS) cold temperature caused by the initiating event is shown in Figure 6. (( }}2(a),(c) The RCS average temperature also decreases as shown in Figure 7. The temperature decrease, and corresponding density increase, causes a reduction in the power detected at the ex-core detectors as shown in Figure 8. (( }}2(a),(c) The overall detected power is shown in Figure 9. (( }}2(a),(c) Cases where detected power in Figure 9 exceeds 115 percent power result in a reactor trip on high power. High power rate reactor trip occurs if the instantaneous detected power in Figure 9 is more than 7.5 percent different than the 30-second-averaged detected power in Figure 10. (( }}2(a),(c) The core power and core power rate of change are shown in Figure 11 and Figure 12, respectively. NuScale Nonproprietary NuScale Nonproprietary

Figure 1: Core Power - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 2: Reactor Coolant System Flow Rate - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 3: Reactor Coolant System Average Temperature - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 4: Reactor Pressure Vessel Lower Plenum Pressure - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 5: Steam Generator Pressure - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 6: Reactor Coolant System Cold Temperature (zoomed-in) - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 7: Reactor Coolant System Average Temperature (zoomed-in) - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 8: Ex-core Detector Decalibration Amount (zoomed-in) - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 9: Ex-Core Detector Power (zoomed-in) - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 10: Ex-Core Detector 30-Second Average Power (zoomed-in) - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 11: Core Power (zoomed-in) - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Figure 12: Core Power Change Rate (zoomed-in) - High Power Rate Reactor Trip Signal Decalibration Sensitivity Study (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Although the sensitivity study was performed for a specific initiating event (( }}2(a),(c), the following general conclusions can be drawn from the sensitivity study and are applicable to other cooldown events:

The decalibration effect on the high power rate reactor trip can be conservatively modeled using a decalibration factor.

Applying a larger decalibration factor is conservative (( }}2(a),(c)

Negative high power rate reactor trips, if they occur, are easy to identify by reviewing plots of output parameters. (( }}2(a),(c) The process for addressing negative high power rate reactor trips is added to TR-0516-49416-P, Revision 4, as shown in the attached markups.

(( }}2(a),(c) the high power rate reactor trip only protects against the types of events it was intended to mitigate; slower reactivity insertions are mitigated by other reactor trips. In addition to these general conclusions, the sensitivity results also confirm the following specific conclusions for the FSAR Section 15.1 analyses:

A decalibration factor of approximately 0.6 percent core power per degree F is reasonable (( }}2(a),(c)

A decalibration factor of approximately 0.6 percent core power per degree F is reasonable (( }}2(a),(c) These specific conclusions regarding the reasonableness of using a decalibration factor of approximately 0.6 percent core power per degree F are applicable to cooldown events ((

}}2(a),(c) Therefore, the treatment of the high power rate reactor trip in EC-104467, Revision 1, and EC-120241, Revision 1, is conservative and appropriate.

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(( }}2(a),(c) appropriate output plots are generated (e.g., Figure 6 through Figure 12) and reviewed to confirm that negative high power rate reactor trips are not occurring. If they are occurring for a cooldown event, (( }}2(a),(c) The process for addressing negative high power rate reactor trips is added to TR-0516-49416-P, Revision 4, as shown in the attached markups. Response to Specific NRC Requests Responses to the specific NRC requests in this RAI are provided by first identifying the NRC request in indented, italicized text, followed by the NuScale response in un-indented, regular text. Revise calculations supporting the decrease in feedwater temperature and increase in steam flow events such that modeled reactor trip signals are consistent between the limiting case and sensitivity studies, unless the sensitivity study is performed to examine modeling of a trip signal. EC-104467, Revision 1, and EC-120241, Revision 1, use modeled reactor trip signals that are consistent among the analyzed cases. The exception is (( }}2(a),(c) where the reactor trip signal modeling is modified specifically to perform a sensitivity study to examine the modeling of the reactor trip signal. EC-104467, Revision 1, and EC-120241, Revision 1, are available in the eRR. If the high power rate trip will be credited in limiting cases, confirm that the modeled change in excore detector signal from a change in downcomer temperature is conservative or accurate by providing, for limiting cases, the magnitude of the modification to the excore detector signal or trip setpoint, trip time, and by showing progressions of downcomer temperature, core average temperature, actual reactor power, and indicated reactor power, and other parameters as necessary such that the phenomena leading to the high power rate trip are clear. NuScale Nonproprietary NuScale Nonproprietary

The high power rate reactor trip (( }}2(a),(c) The justification in this response, including sensitivity study results, demonstrate that the high power rate reactor trip treatment is reasonable and conservative. The sensitivity results provided in this response show the magnitude of the modification to the excore detector signal, trip time, downcomer temperature, core average temperature, actual reactor power, and indicated reactor power. These parameters clearly demonstrate the phenomena leading to the high power rate reactor trip (( }}2(a),(c) Similar figures are generated for each case in the analyses for EC-104467, Revision 1, and EC-120241, Revision 1. The figures are provided for limiting cases in EC-104467, Revision 1, and EC-120241, Revision 1. Additionally, update portions of the licensing basis discussing the methodology for evaluation of overcooling transients (i.e., the non-LOCA LTR or a subsection of FSAR 15.0) to specify how the evaluation model accounts for changes in excore detector signal caused by changes in downcomer temperature when assessing high power rate trips. FSAR Table 15.0-7 contains a footnote to describe how the effects of downcomer temperature are accounted for in the high power reactor trip signal for cooldown events. FSAR Table 15.0-7 is revised as shown in the attached markup to apply this same footnote to the high power rate reactor trip signal and expand the footnote to also describe how the effects of downcomer temperature are accounted for in the high power reactor trip signal for cooldown events. In the response to audit question A-15.1.1-7, NuScale provided markups of TR-0516-49416-P, Revision 4, Non-Loss-of-Coolant Accident Analysis Methodology, to expand the discussion of the treatment of power-related reactor trip signals during cooldown events. The discussion described the treatment for the high power rate reactor trip signal in addition to the high power reactor trip signal. (( }}2(a),(c) NuScale has reviewed the TR-0516-49416-P, Revision 4, markups previously provided and concludes that they are accurate and consistent with this response, with the exception that the NuScale Nonproprietary NuScale Nonproprietary

process for addressing negative high power rate reactor trips is added to TR-0516-49416-P, Revision 4, as shown in the attached markups. Provide justification that this methodology results in a conservative or accurate trip time. If the magnitude of this effect is biased in a specific direction, provide justification that trip time will be accurate or conservative in circumstances in which the effect encourages a high power rate trip (e.g., downcomer temperature is increasing when a positive rate trip is assessed) and discourages a high power rate trip (e.g., downcomer temperature is decreasing when a positive rate trip is assessed), as information provided during the audit does not address all of these circumstances. This response provides physics-based justifications, supplemented by sensitivity results, that the treatment of the high power rate reactor trip signal is conservative for cooldown events. (( }}2(a),(c) If justification for this biasing is based on a specific event progression or timing of the high power rate trip, state this basis such that conservatism or appropriateness of the approach for overcooling events is clear. The physics-based justifications are not event-specific or specific to a particular actuation timing. The sensitivity study is performed for a specific event (( }}2(a),(c) but the generic conclusions are applicable to cooldown events in general as described above. (( }}2(a),(c) Therefore, discussion of the sensitivity study, or its limitations, is not required to be added to TR-0516-49416-P, Revision 4. Specify any modeling details (e.g., method of modeling this effect such as a scalar multiplication on core power, formulas used to assess the high power rate trip) needed to support justification that the specific method is conservative, as information provided during the audit does not sufficiently specify this information. NuScale Nonproprietary NuScale Nonproprietary

The methodology in TR-0516-49416-P, Revision 4, including markups identified in the previous response to audit question A-15.1.1-7 and those provided with this response, identifies how the effects are modeled: the high power reactor trip setpoint is effectively increased and a decalibration factor (scalar multiplier) is applied for the inputs to the high power rate reactor trip signal. The implementation of this methodology is documented in the FSAR, where Table 15.0-7, including markups provided with this response, repeats this approach while also describing the specific values used. The event-specific analysis output provides confirmation that the results of the implemented methodology are conservative. Impact on US460 SDAA: FSAR Section 15.0 and Topical Report TR-0516-49416, Non-Loss-of-Coolant Accident Analysis Methodology, have been revised as described in the response above and as shown in the markups provided in this response. NuScale Nonproprietary NuScale Nonproprietary

NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-51 Draft Revision 2 Audit Question A-15.1.1-3 RAI 10357 Question 15.1.1-7 Table 15.0-7: Analytical Limits and Time Delays Signal(1) Analytical Limit Basis and Event Type Actuation Delay High Power 115%(2) RTP ( 15% RTP) 25% RTP (<15% RTP) This signal is designed to protect against exceeding CHF limits for reactivity and overcooling events. 2.0 sec Source and Intermediate Range Log Power Rate 3 decades/min This signal is designed to protect against exceeding CHF and energy deposition limits during startup power excursions. Variable High Power Rate +/-7.5%(2) RTP/30 sec This signal is designed to protect against exceeding CHF limits for reactivity and overcooling events. 2.0 sec High Source Range Count Rate 5.0 E+05 counts per second(3) This signal is designed to protect against exceeding CHF and energy deposition limits during rapid startup power excursions. 3.0 sec High Subcritical Multiplication 3.2 This signal is designed to detect and mitigate inadvertent subcritical boron dilutions in operating Modes 2 and 3. 150.0 sec High RCS Hot Temperature 620°F This signal is designed to protect against exceeding CHF limits for reactivity and heatup events. 8.0 sec High RCS Average Temperature 555°F This signal is designed to protect against exceeding CHF limits for reactivity events. 8.0 sec High Containment Pressure 9.5 psia This signal is designed to detect and mitigate RCS or secondary leaks above the allowable limits to protect RCS inventory and ECCS function during these events. 2.0 sec High Pressurizer Pressure 2100 psia This signal is designed to protect against exceeding RPV pressure limits for reactivity and heatup events. 2.0 sec High Pressurizer Level 80% This signal is designed to detect and mitigate CVCS malfunctions to protect against overfilling the pressurizer. 3.0 sec Low Pressurizer Pressure 1850 psia(4) This signal is designed to detect and mitigate high-energy line break (HELB) events from the pressurizer vapor space and protect RCS subcooled margin for protection against instability events. 2.0 sec Low-Low Pressurizer Pressure 1200 psia(5) This signal is designed to protect RCS subcooled margin for protection against instability events. 2.0 sec Low Pressurizer Level 35% This signal is designed to detect and mitigate pipe breaks to protect RCS inventory and ECCS functionality during LOCAs, primary HELB outside containment events, or SGTF, and to protect the pressurizer heaters from uncovering and overheating during decrease in RCS inventory events. 3.0 sec Low-Low Pressurizer Level 15% This signal is designed to detect and mitigate pipe breaks to protect RCS inventory and ECCS functionality during LOCAs, primary HELB outside containment events, or SGTF. 3.0 sec

NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-53 Draft Revision 2 High Under-the-Bioshield Temperature 250°F This signal is designed to detect high energy leaks or breaks at the top of the NPM under the bioshield to reduce the consequences of HELBs on the safety-related equipment located on top of the module. 8.0 sec Notes:

1. Interlocks, permissives, and overrides for these signals are described in Table 7.1-5.

2. The overcooling event analyses account for decreased power detectionuncertainty due to decreasing downcomer temperature. For the high power signal, the detected powerThe reductionuncertainty is 7% RTP for downcomer temperature decreases up to 10°F and the reduction is scaled upwards from 7% by approximately 0.6% core power per degree F for downcomer temperature decreases beyond 10°F. For the high power rate signal, the detected power reduction uses the same scalar factor of approximately 0.6% core power per degree F for both downcomer temperature decreases up to 10°F and for downcomer temperature decreases beyond 10°F.

3. The high count rate trip is treated as a source range over power trip that occurs at a core power analytical limit of 500 kW, which functionally equates neutron monitoring system counts per second to core power in watts. This trip is bypassed once the intermediate range signal is established.

4. If RCS hot temperature is above 500°F as shown in Figure 4.4-2.

5. If RCS hot temperature is below 500°F as shown in Figure 4.4-2.

6. If RCS hot temperature is above 500°F.

7. The RPV water level is presented in terms of elevation where reference zero is the bottom of the module assembly (at the bottom of the reactor pool). The range accommodates instrumentation uncertainty.

8. Normal AC voltage is monitored at the battery chargers for the EDAS. The analytical limit is based on an average voltage below 80% of normal.

Table 15.0-7: Analytical Limits and Time Delays (Continued) Signal(1) Analytical Limit Basis and Event Type Actuation Delay

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 534 Closure of the FWIVs isolates the SGs from the colder feedwater, 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 terminates with the NPM in a safe, stable condition. Table 7-5 lists the relevant acceptance criteria, SAF, and LOP scenarios. Audit Question A-15.1.1-7 The limiting MCHFR typically occurs when the event is initiated from full power conditions. (( }}2(a),(c) For overcooling events, the high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. (( }}2(a),(c) Audit Question A-15.1.1-7 RAI 10357 Question 15.1.1-7 The increase in downcomer density from the cooldown causes a decrease in neutron flux at the excore neutron detectors associated with power-related reactor trip signals in the NPM. Therefore, the power used to generate power-related reactor trips accounts for this decalibration effect if the power-related reactor trips are credited in the analysis. For the high power reactor trip, the high power analytical limit is effectively increased to account for the decalibration effect, for example from 120 percent (Table 7-3) to 125 percent RTP. For the high power rate reactor trip, accounting for the decalibration effect delays actuation of the trip due to the time for the excore detectors to register the power increase. (( }}2(a),(c) Neglecting reactor trip on a power-related reactor trip signal is a

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) Additional sensitivity studies are performed on other 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, 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 544 Audit Question A-15.1.1-7 A rapid (step) increase in feedwater flow is simulated. The limiting MCHFR typically occurs when the event is initiated from full power conditions, (( }}2(a),(c) For overcooling events, the high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. The increase is based on an appropriate decalibration factor (change-in-power-per-change-in-temperature) and considering the downcomer temperature decrease during the overcooling events. Audit Question A-15.1.1-7 RAI 10357 Question 15.1.1-7 The increase in downcomer density from the cooldown causes a decrease in neutron flux at the excore neutron detectors associated with power-related reactor trip signals in the NPM. Therefore, the power used to generate power-related reactor trips accounts for this decalibration effect if the power-related reactor trips are credited in the analysis. For the high power reactor trip, the high power analytical limit is effectively increased to account for the decalibration effect, for example from 120 percent (Table 7-3) to 125 percent RTP. For the high power rate reactor trip, accounting for the decalibration effect delays actuation of the trip due to the time for the excore detectors to register the power increase. (( }}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.

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 555 RTS and isolation of the secondary system on the low steam pressure or low steam superheat signal. Isolation of the secondary system, if it has not already occurred, and actuation of DHRS occur post-trip on other MPS signals. Closure of the FWIVs and MSIVs isolates the steam generators, which ends 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 terminated with the NPM in a safe, stable condition. Table 7-17 lists the relevant acceptance criteria, SAF, and LOP scenarios. Audit Question A-15.1.1-7 The limiting MCHFR typically occurs when the event is initiated from full power conditions, and the increase in steam flow is such that the immediate RTS on low steam pressure is avoided and RTS is actuated sometime later (after the minimum CHFR conditions develop). For overcooling events, the high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. The increase is based on an appropriate decalibration factor (change-in-power-per-change-in-temperature) and considering the downcomer temperature decrease during the overcooling events. Audit Question A-15.1.1-7 RAI 10357 Question 15.1.1-7 The increase in downcomer density from the cooldown causes a decrease in neutron flux at the excore neutron detectors associated with power-related reactor trip signals in the NPM. Therefore, the power used to generate power-related reactor trips accounts for this decalibration effect if the power-related reactor trips are credited in the analysis. For the high power reactor trip, the high power analytical limit is effectively increased to account for the decalibration effect, for example from 120 percent (Table 7-3) to 125 percent RTP. For the high power rate reactor trip, accounting for the decalibration effect delays actuation of the trip due to the time for the excore detectors to register the power increase. (( }}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. 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 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 562 7.2.4.1 General Event Description The steam line break event ranges from small breaks to double ended ruptures of a main steam line causing an increase in steam flow and an over cooling of the RCS. This event can occur inside or outside the containment vessel (CNV). A break inside CNV causes a rapid pressurization of the CNV resulting in a reactor trip and CNV isolation with a DHRS actuation. This break location is non-limiting for pressure and MCHFR but potentially challenging to the DHRS as one loop is disabled with the break inside the CNV. A steam line break outside of the CNV causes an increase in steam flow event that causes either a low SG pressure trip or a high core power trip due to the reactor power response from the decreased RCS temperature. The break flow is stopped by the MSIVs closing and depressurization of the steam system piping. Smaller breaks cause a slower loss of secondary pressure due to the increased steam demand that could cause a high core power trip. These smaller breaks can result in a significant delay in detection time, making the small break cases potentially challenging for MCHFR. Reactor trip and transition to stable DHRS flow eventually terminate the transients and bring the NPM to a safe, stable condition. Audit Question A-15.1.1-7 RAI 10357 Question 15.1.1-7 The increase in downcomer density from the cooldown causes a decrease in neutron flux at the excore neutron detectors associated with power-related reactor trip signals in the NPM. Therefore, the power used to generate power-related reactor trips accounts for this decalibration effect if the power-related reactor trips are credited in the analysis. For the high power reactor trip, the high power analytical limit is effectively increased to account for the decalibration effect, for example from 120 percent (Table 7-3) to 125 percent RTP. For the high power rate reactor trip, accounting for the decalibration effect delays actuation of the trip due to the time for the excore detectors to register the power increase. (( }}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.For this overcooling event, the

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 563 high power analytical limit is increased, for example from 120 percent (Table 7-3) to 125 percent RTP. This increase accounts for the decalibration of the excore neutron detectors as downcomer density increases in response to a cooldown event. The increase is based on an appropriate decalibration factor (change-in-power-per-change-in-temperature) and considering the downcomer temperature decrease during the overcooling events. Audit Question A-NonLOCA.LTR-50 (( }}2(a),(c) The relevant acceptance criteria, SAF, and LOP scenarios are listed in Table 7-22. Audit Question A-NonLOCA.LTR-66 Table 7-22 Acceptance criteria, single active failure, loss of power scenarios - steam line break Acceptance Criteria / Single Active Failure / Loss of Power Scenarios of Interest Discussion MCHFR Radiological consequences Critical heat flux is potentially challenged for this overcooling event. (Reactivity insertion rates from the overcooling event are insufficient to challenge fuel centerline temperature.) A postulated break in the main steam line is evaluated for radiological consequences. Failure of one MSIV to close on the train with break Failure of one FWIV to close on the train with the break MSIV single failure typically has no effect on MCHFR since DHRS actuation is not before the time when MCHFR occurs. MSIV single failure is typically limitingpotentially bounding for mass releases for break locations outside the CNV if mass release is calculated for use in the downstream radiological analysis. FWIV single failure typically has no effect on MCHFR since DHRS actuation is not before the time when MCHFR occurs. FWIV single failure is typically limitingpotentially bounding for mass releases for break locations inside the CNV if mass release is calculated for use in the downstream radiological analysis. No loss of power Loss of power scenarios typically terminate feedwater and/or trip the reactor, thus mitigating the overcooling event.

RAIO-177177 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-177178

AF-177178 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. 10357 R1, Question 15.1.1-7) 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 10357 R1, Question 15.1.1-7. 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-177178 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 December 18, 2024. Mark W. Shaver}}