ML26026A316

NRC Staff Responses to Commission Hearing Questions

ML26026A316
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	01/26/2026
From:	Julie Ezell NRC/OGC
To:	NRC/OCM
SECY RAS
References
RAS 57589, 50-613-CP
Download: ML26026A316 (0)

Text

January 26, 2026 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the Matter of US SFR OWNER, LLC (Kemmerer Power Station, Unit 1)

Docket No. 50-613-CP NRC STAFF RESPONSES TO COMMISSION HEARING QUESTIONS Pursuant to the Commissions Order (Transmitting Hearing Questions) dated January 12, 2026, the U.S. Nuclear Regulatory Commission (NRC) staff (Staff) hereby responds to the questions posed in that Order. These questions generally pertain to the NRC Staffs safety review,1 review process, or final environmental impact statement.2 The Commissions Order directed some questions only to the NRC Staff and some to both the Staff and the Applicant. The attachment to this filing presents the NRC Staffs responses to the questions directed to the Staff.

Respectfully submitted,

/Signed (electronically) by/

Julie G. Ezell Counsel for NRC Staff Mail Stop: O15-B04 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Telephone: (301) 287-9157 E-mail: julie.ezell@nrc.gov 1 Safety Evaluation Related to the U.S. SFR Owner, LLC Construction Permit Application for the Kemmerer Power Station Unit 1 (November 2025) (ADAMS Accession No. ML25329A252).

2 Environmental Impact Statement for the Construction Permit Application for Kemmerer Power Station Unit 1 (Final Report) (October 2025) (ML25287A017).

1 NRC STAFF RESPONSES TO COMMISSION HEARING QUESTIONS

1. The Staffs Information Paper notes that the Staff implemented an expedited and risk-informed legal review of this application.

a. Discuss how the approach to the legal review for this application differs from previous construction permit or combined license applications.

b. Describe how the Staff consulted with the Office of the General Counsel in determining the risk levels posed by different areas of review and what level of legal review would be given to various areas or chapters of the safety evaluation (SE) and the environmental review.

c. Has the Staff identified any efficiencies or challenges with this approach at this time?

d. Does the Staff intend to continue a limited legal review approach at the operating license stage?

Staff Response: After establishing the schedule for this review, in consideration of the direction in the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy (ADVANCE) Act of 2025 and Executive Order 14300, Ordering the Reform of the Nuclear Regulatory Commission, to expedite reviews of reactor applications, the Staff accelerated its review of the Kemmerer Power Station Unit 1 (KU1) construction permit (CP) application by 9 months, which shortened managements, OGCs, and the ACRSs review time to approximately 3 months and allowed for a parallel review as the Staff completed other portions of the SE.1 In previous reviews, the concurrence review process occurred sequentially.

To provide the maximum time for the technical staff to conduct its review, the Staff streamlined the concurrence review process and made decisions about the scope of requested legal review, including which chapters or sections of the SE would be reviewed and the type of legal review requested. When determining the portions of the SE for which Staff would request OGC review, the Staff considered a variety of factors, including whether the applicant proposed new approaches to meet regulatory requirements, whether the facilitys design raised questions about alignment with the existing regulatory framework, and whether the applicant was seeking exemptions. As a result, the Staff adjusted the scope (i.e., review areas and level) of the requested OGC review from that requested for previous construction permit or combined license applications. The Staff did not amend the review approach for the EIS (there was a full legal review), but the Staff may implement an abbreviated approach for the legal review of future environmental documents.

The Staff plans on using this approach for the KU1 operating license review and all other future advanced reactor reviews. For future reviews, the Staff intends to work with OGC proactively to identify areas for review, which will be important for resource planning and could result in additional efficiency improvements.

1 The NRC Staff notes that consistent with the agencys Office of Management and Budget approved lapse plan, development of this SE continued during the lapse in appropriations from October 1 to November 13, 2025, which included OGC review. OGC and the Staff worked during the shutdown to complete the Kemmerer review on time. These unique circumstances would need to be considered in using this approach in the future.

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2. In SECY-25-0102, the Staff identified areas of the review that it believed posed the greatest risk and requested OGC review of those aspects of the Staffs review.

What were the areas of risk that were identified and how were those addressed in the construction permit (CP)?

Staff Response: The Staff considered as higher risk areas where the applicant proposed new or novel approaches to meeting regulatory requirements, sought exemptions from the regulations, or the facilitys design raised questions about alignment with the existing regulatory framework. The Staff considered as lower risk areas where the application was consistent with long-standing regulatory practices. The Staff adjusted its requests for legal review based upon these risk considerations and considered OGCs advice in the development of its final safety evaluation. OGCs advice was instrumental in the Staffs navigation of complex regulatory issues.

3. USO has elected to use the Licensing Modernization Project (LMP) method, as described in NEI 18-04, Rev. 1, and corresponding use of a probabilistic risk assessment (PRA) to support the licensing basis development. In reviewing the application of the LMP method, the Staff has generally found that the information provided adequately supports the issuance of a construction permit; however, there are many areas identified that would need to be reevaluated at the operating license application stage. For example, there are uses of conservative assumptions to select mechanistic source terms for some event sequences, rather than developing event specific mechanistic source terms. Further, while a hazards screening was performed based on the Staffs guidance on PRA acceptability in Regulatory Guide (RG) 1.247, the Staffs SE stated that USO will revise the justifications and update the PRA self-assessment to support the operating license (OL) application. As the safety and licensing bases analyses are refined and revised, the outcomes of the LMP process, including the safety classification designations, could change. Systems performing safety-related functions (e.g., Reactor Air Cooling (RAC)) and safety-related barriers serving as functional containment against radionuclide releases (e.g., Ex-Vessel Handling Machine (EVHM) cask barrier and Pin Removal Cell (PRC) hot cell barrier) could be classified at lower safety classification levels. On the other hand, SSCs that are currently classified as either non-safety related or non-safety related with special treatment (NSRST) can be reclassified as higher safety classification levels, including being identified as safety-related. Such changes to SSC classifications are expected to impact their quality assurance and construction.

a. What is the process for reclassification, including identification and finalization, of structures, systems, and components during the construction phase?

Staff Response: SSC safety classification changes between any CP and the OL application are anticipated under the 10 CFR Part 50 two-step licensing process and are not unique to applications following the LMP process. However, LMP provides for the incorporation of progressively more detailed design information in a structured and iterative manner, ultimately

3 supporting a final set of SSC classifications prior to the OL stage.2 If new information results in a change to an SSC classification, the applicant must address the change in the OL application.

Changes that significantly affect the Staffs findings in the CP SE may necessitate a permit amendment.

Given the preliminary nature of SSC classifications at the CP stage, the Staffs review focused, in part, on the process the applicant used to classify SSCs under LMP. A detailed discussion of the applicants SSC classification process is provided in Chapter 5 of the SE with supporting information on licensing basis event (LBE) selection and defense-in-depth (DID) evaluation inputs in Chapters 3 and 4 of the SE, respectively. The Staff also conducted an audit of the applicants processes, including internal guides and procedures, related to SSC classification, as discussed in Chapter 5 of the SE. Based on its review, the staff concluded that if there were changes to the major features or SSCs, the process described in the application would generally be able to address those changes and account for their effects.

b. How will potential impacts from the refinement of safety and licensing basis information on the principal architectural and engineering criteria for the design or major features and components that are identified for protection of the health and safety of the public in accordance with the regulations of 10 C.F.R. § 50.35 be reflected in the licensing bases and construction activities?

Staff Response: The Staff plans to maintain awareness of changes to SSC classification and design during the construction phase through review of the reports required by the proposed permit conditions and the construction oversight process. The applicant can manage risks to the OL review by proactively engaging with the Staff when issues, including substantive SSC classification and design changes, arise during construction. More significant changes may necessitate a permit amendment. The Staff notes that the applicant submitted a regulatory engagement plan that includes planned robust pre-application engagement on numerous topics prior to the OL application submittal (ML25273A340).

The requirement to provide the principal architectural and engineering criteria of the design specified in 10 CFR 50.35 is met, in part, through the Kemmerer Unit 1 principal design criteria (PDC). These PDC are provided in the preliminary safety analysis report (PSAR), which incorporates by reference the NRC Staff-approved topical report NATD-LIC-RPRT-0002-A, Rev.

1, Principal Design Criteria for the Natrium Advanced Reactor (ML24283A066). Substantive changes to the PDC would represent a shift in the applicants approach to demonstrate the safety of the facility, and would need to be reflected by significant changes to the plant design bases as documented in the PSAR (particularly aspects documented in Chapters 1, 3, 4, 5, 6, and 7 of the PSAR). The applicant would have to evaluate such changes to identify whether a permit amendment is needed.

Similarly, substantial changes to the major features and components identified for the protection of the health and safety of the public (e.g., SR SSCs and select NSRST SSCs) would also 2 For example, NEI 18-04 Section 3.2.3, Evolution of LBEs [Licensing Basis Events] Through Design and Licensing Stages, discusses how the LBE selection process is applied iteratively as the design evolves (as shown in NEI 18-04 Figure 3-2). This process includes SSC safety classification and DID adequacy evaluation steps.

4 represent a shift in the applicants approach. In the Staffs expert opinion, the design of these SSCs is comparatively mature both in design and analysis, with areas of greater uncertainty in the design managed, in part, through R&D items (e.g., the heat transfer associated with the reactor air cooling system). Changes to the functional design of these SSCs would have significant effects on their design bases described in the PSAR and would affect other plant SSCs and their design bases due to the integrated nature of developing an application using LMP. Thus, although there could be changes to the safety classification and design of certain SSCs, it is unlikely that there would be substantial changes to the major features and components.

Though major changes are not expected, as noted in the question, refinement of the analysis or design changes could result in changes to SSC safety classifications, which could then result in changes to the quality assurance and construction of the subject SSCs. NEI 18-04 discusses SSC special treatment in the LMP process. Quality assurance and use of design and construction codes and standards beyond normal industrial practices are examples of SSC special treatments in the LMP process. Chapters 6, 7, and 8 of the PSAR explain the design, construction, and quality assurance standards applied to all safety-related (SR) SSCs (e.g., use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III for SR primary coolant boundary components), which the Staff evaluated in the corresponding chapters of the SE. However, the Staffs review of special treatments for NSRST SSCs at the CP was conducted on an SSC-specific basis, because the special treatments vary depending on the safety-significance of a given SSC and the functions it performs. Due to the graded treatment of NSRST SSCs, new SSCs that are identified by the applicant as NSRST will be an area of focus during the OL application review.

4. The Staff notes two deviations from the NEI 18-04 / RG 1.233 licensing basis event (LBE) methodology (SER pages 3-74 to 3-77). Explain the deviations in plain English and their impacts to the CP and a subsequent OL review.

Staff Response:

First Deviation - Categorization of Beyond Design Basis Events Description of the Deviation The first deviation discussed in the Staffs SE relates to how the applicant categorized beyond design basis events (BDBEs).

In the Licensing Modernization Project (LMP) process, an applicant identifies a set of licensing basis events (LBEs)3 and estimates their likelihood (i.e., frequency) and consequences. Based on the likelihood, these events are grouped into categories: anticipated operational occurrences (AOO), design basis events (DBE), and BDBEs. Each event is then evaluated against the frequency-consequence (F-C) target curve provided in NEI 18-04. The LBE categorization and evaluation are used, in part, to determine the functions and the safety classifications of 3 LBEs are defined in NEI 18-04 as the entire collection of event sequences considered in the design and licensing basis of the plant. An applicant develops the set of LBEs for a facility following the process provided in Section 3 of NEI 18-04 by iterating an initial set of proposed events with insights from probabilistic risk assessment.

5 structures, systems, and components (SSCs) that are needed to protect public health and safety.

Each LBE has a range of possible frequencies, based on its uncertainty. An LBE is categorized as an AOO or DBE if its mean frequency is within the range specified in the guidance. However, to categorize an LBE as a BDBE, RG 1.233 indicates that the upper end of the LBE frequency range should be compared to the lower end of the BDBE range in NEI 18-04. This ensures that some less likely events that are more uncertain are appropriately categorized as BDBEs.

However, the applicant deviated from the guidance by comparing the mean LBE frequency (instead of the upper end of the range) to the lower end of the BDBE frequency range.

Impact to the Construction Permit (CP) Application Review The applicants deviation caused two events to not be included as BDBEs in the PSAR. Details on the frequency and consequence of these two events were provided by the applicant in response to NRC Request for Confirmation of Information (RCI)-1 (ML25259A180) and were included in Table 3.4-1 of the SE.

The impact of this deviation on the CP application review was minimal because, though the applicant used an alternative criterion to categorize the LBEs, the applicant conducted the appropriate evaluation of the LBEs against the F-C target. The applicant was able to conduct the appropriate evaluation despite the alternative criterion because the categorization was not used to determine which LBEs were evaluated against the F-C curve. The applicant then classified the equipment needed to mitigate these two events as safety-related (SR). Based on the fact that the applicants process appropriately evaluated events against the F-C curve, the Staff concludes that the effect of this deviation was not significant and the applicants approach ultimately led to the appropriate identification of safety functions and SSC safety classifications at the CP stage.

Impact to a Subsequent Operating License (OL) Application Review If the applicant continues to use alternative criteria to categorize events as BDBEs in development of the OL application, as indicated in its response to RCI-1, there could be impacts for the OL application review. The Staff expects these impacts to be similar to the impacts for the CP application review and will approach them in the same manner.

Second Deviation - Selection of Design Basis Accidents Based on Uncertainty Bands Description of the Deviation The second deviation discussed in the Staffs SE is related to how the set of design basis accidents (DBAs) was selected.

In the LMP process, a DBA is derived from each DBE by evaluating the event assuming only safety-related SSCs are available.4 Following this process, one DBA may correspond to multiple DBEs. Each DBA is then analyzed using conservative, deterministic safety analyses, which demonstrate the performance of the safety-related SSCs and ensure that they are adequate to meet the 10 CFR 50.34 dose criteria when only they are available.

4 As stated in NEI 18-04, DBEs may credit non-safety-related SSCs, while DBAs credit only safety-related SSCs.

6 As discussed earlier in this response, the LMP process includes uncertainty on the LBE frequency estimate. While categorization typically relies on the mean value, the uncertainties in the LBE frequency may extend into neighboring categories. For example, the upper end of a BDBEs frequency range may extend into the DBE frequency region.

NEI 18-04 indicates that if a specified uncertainty range for an LBE straddles an LBE categorization frequency boundary, the LBE should be evaluated in both categories. As discussed further on page 3-75 of the SE, DBEs that are evaluated because the LBE uncertainties extend into the DBE frequency region should have associated DBAs. In the CP application, USO identified a number of LBEs with uncertainties that extended into the DBE region without associated DBAs.

Impact to the CP Application Review The LBEs in question include six AOOs and eleven BDBEs whose lower and upper uncertainty ranges, respectively, extend into the DBE region. The Staff evaluated these events to determine whether the information provided in the CP application was sufficient to demonstrate that safety-related SSCs can perform the necessary safety functions and that the overall set of safety-related SSCs is sufficient to ensure the 10 CFR 50.34 dose criteria can be met.

As discussed in Sections 3.5 and 3.7 of the SE, the Staff found that several of these LBEs were appropriately bounded by existing DBAs or other LBEs that only credited safety-related SSCs.

In other cases, the LBEs themselves relied only on safety-related SSCs and demonstrated doses low enough that the Staff concluded that a conservatively-evaluated DBA would be unlikely to exceed the 10 CFR 50.34 dose criteria. In the Staffs engineering judgement, identifying DBAs for these LBEs would not have resulted in changes to SSC safety classifications.

The Staff identified two events where a new DBA identified at the CP stage would have potentially affected the application. The first of these events, discussed in Section 3.7.1.1 of the Staffs SE, was a BDBE that relies on a non-safety-related with special treatment alternative shunt trip, which was not yet fully integrated into the CP application.5 The Staff concluded in the SE that a review of the DBA evaluation associated with this BDBE can reasonably be left to the OL stage based on the preliminary nature of the alternative shunt trip design, the design criteria for safety-related and non-safety-related with special treatment control systems, and the applicants design control processes. The Staff discussed the implications of a DBA in the final design for this event, including the potential to reclassify the alternative shunt trip as safety-related, during the audit.

The second event, discussed in Section 3.7.1.5 of the Staffs SE, was a BDBE involving the pool immersion cell (PIC). While the applicant analyzed the PIC for the CP application, this analysis was preliminary. Since the CP application was submitted, the PIC design was updated, in part due to this deviation, to incorporate several additional preventative controls that are anticipated to reduce the frequency of events involving the PIC. Specifically, in a supplement dated September 17, 2025 (ML25260A002), the applicant described its plans for revisions to the design and analysis for the OL stage, including updated events and event-specific source terms, to better represent the configuration of the PIC during transients. Based on the applicants plans 5 See note 1 to PSAR Table 5.2-4.

7 for design changes and reanalysis, the staff determined it was reasonable to leave further consideration of events involving the PIC to the OL stage.

Impact to a Subsequent OL Application Review For the OL application, the applicant will develop DBAs for LBEs with uncertainties that cross into the DBE range, as stated in PSAR section 3.9. The Staff intends to evaluate any new DBAs in the same manner as the DBAs were evaluated for the CP application.

5. The Staff notes that fuel handling events represent the greatest risk for radionuclide release. Besides fuel handling events, which KU1 LBEs are closest to or most significant to the Frequency-Consequence target line? For these events, please explain the key uncertainties and how information in the CP would inform the subsequent OL review.

Staff Response: All the LBEs identified as risk-significant6 in the PSAR are fuel handling events. A list of the frequencies and consequences for all LBEs is in the applicants response to Request for Confirmation of Information (RCI)-3 (ML25259A180). No other events approach the threshold for risk significance under NEI 18-04, even when considering events based on the 95th percentile of their frequency or consequence uncertainty.7 The staff examined other types of events to identify those that are closest to the F-C target curve. The next most significant events are design basis events (DBEs) representing primary sodium processing system (SPS) leaks, beyond design basis events (BDBEs) representing certain loss of offsite power (LOOP) scenarios, and BDBEs for core blockages and local faults.

Note that these events have substantial margin to the F-C target curve.8 The key uncertainties for these events generally relate to SSC performance and reliability and considerations related to mechanistic source term analyses. The information in the CP application will enable the Staff to focus its review of the OL application on the most significant aspects of the relevant LBE.

The primary SPS leak events are initiated by leaks from the primary SPS into the surrounding cell. The analyses are discussed in Section 3.6.1.3 of the safety evaluation (SE), which provides the key characteristics of the event, including the volume of the sodium leaked and the performance of the barrier surrounding the SPS, known as the SPS cell barrier. Because the dominant dose contributor for these events is sodium-24, which has a short half-life, the timing of the scenario and any hold-up in the cell plays a key role in the consequence analysis. At the CP stage, the SPS cell barrier performance did not explicitly account for connected systems to provide an inert environment and heating, ventilation and air conditioning (HVAC) for the cell, and timing of releases from the cell to the environment were conservatively assumed. At the OL stage, the staff expects to focus on the SPS pump trip and isolation valve closures, which contribute to the volume of sodium that can leak into the cell, and the performance of the SPS cell barrier.

6 Defined in NEI 18-04 as LBEs with frequences within 1% of the frequency-consequence (F-C) target and with site boundary doses exceeding 2.5 mrem over thirty days.

7 PSAR Tables 4.2-1 and 4.2-2 provide lists of the risk-significant LBEs as identified based on the frequency and consequence mean and 95th percentiles, respectively.

8 The DBEs have estimated doses of less than 0.1 rem (10% of the lowest consequence on the F-C target curve for DBEs), while the BDBEs have estimated doses of less than 1 rem (4% of the lowest consequence on the F-C target curve for BDBEs).

8 In the LOOP BDBEs, a loss of offsite power results in a loss of primary coolant flow due to the loss of power to the primary sodium pumps (PSPs). As described in Section 3.7.1.1 of the SE, the normal scram system fails to function as intended and control rod insertion is provided by the scram motor drive-in function or the alternative shunt trip.9 These events result in fuel failure due to the mismatch in power and flow as the PSPs coast down at the same time as the control rods are inserted. Consequences are driven by the characteristics of the radionuclide transport through the primary sodium and the functional containment boundary performance. The frequency of these events is driven by the reliability of the normal scram systems, including the scram valves and the reactor protection system, as well as the reliability of the scram motor drive-in and alternative shunt trip functions. At the OL stage, the staff expects to focus on the PSP coast down characteristics, the timing of the alternative shunt trip and scram motor drive-in and resulting control rod insertion curves, and the source term and radionuclide transport calculations.

The core blockage and local fault events are based on the assumed failure of an entire fuel assembly while at power resulting from a fuel assembly blockage, misload, or other issue that could affect a single fuel assembly. These events are described in Section 3.7.1.6 of the SE.

Because a full fuel assembly is assumed to fail in these events in order to bound a range of possible scenarios, the results are most dependent on the radionuclide release from the fuel, radionuclide transport through the primary coolant, and performance of the functional containment barriers. At the OL stage, the staff expects to focus on these characteristics, as well as any manual actions to isolate the sodium cover gas system or scram the reactor.

6. The Natrium sodium fast reactor design considers the needs for protection from sodium fires that could result from liquid sodium interactions with air, concrete, and water. Section 8.2 of PSAR describes the Applicants approach for fire protection. This includes following the guidance in RG 1.189 in the fire protection design strategy and performance of a fire PRA using NUREG/CR-6850, which the Staff will evaluate as part of the operating license application. At the preliminary design stage, the risk significance has not yet been determined for sodium fire events and the structures, systems, and components used for prevention and mitigation.

Considering the potential for significant sodium fires that could progress to result in radionuclide releases, and noting that RG 1.189 and NUREG/CR-6850 have been developed for LWR plants, have the staff and applicant assessed the availability of design standards for detection, suppression, and support systems that would be applicable to Kemmerer Power Station? Given that fire protection features may impact the design and construction of SSCs, how will development of the fire protection strategy be coordinated with construction activities?

Staff Response: Although RG 1.189 was developed for light-water reactors, it states that non-light-water reactor designs should meet the overall fire protection objectives and guidance in the RG as they relate to safe shutdown and radiological release.10 In addition to following RG 1.189 9 The Staff notes SE page 3-111 contains a typographical error and refers to the anticipatory automatic seismic trip system but should have referred to the alternative shunt trip.

10 RG 1.189, Revision 5, Section C.8.5.

9 and NUREG/CR-6850, USO states in Section 7.5.2.3.4 of the PSAR that the plant design incorporates features consistent with ANS-54.8-1988, Liquid Metal Fire Protection in LMR

[liquid metal reactor] Plants. While this standard was withdrawn in July 2000, it was subsequently re-issued in 2025, with minimal changes, as ANS-54.8-2025.11 ANS-54.8 provides guidelines and minimum design requirements for preventing, detecting, and suppressing liquid metal fires and mitigating their consequences in liquid metal reactors, including sodium fast reactors.12 The Staff is currently participating in a working group that intends to update ANS-54.8-2025 to incorporate risk-informed, performance-based approaches and updated references. After the ANS 54.8-2025 standard is updated, the staff plans to consider it for endorsement.

ANS-54.8 provides standards for those aspects of fire prevention and protection unique to liquid metal reactors that are not covered in other codes and standards. For the design of fire detection and conventional fire suppression systems, ANS 54.8 references other standards, such as National Fire Protection Association (NFPA) 10, Standard for Portable Fire Extinguishers, NFPA 70, National Electrical Code (NEC), and NFPA 72, National Fire Alarm and Signaling Code. However, for the design of suppression and support systems unique to sodium, such as catch pans and suppression decks, there are no standards specific to sodium fire hazards. The Staff expects USO to design these SSCs to applicable mechanical, electrical, or structural codes and standards based on design loadings and conditions. PSAR Table 1.4-5 lists the codes and standards used in the design, including IEEE standards and ASME standards. The Staff will review the final design of these SSCs during the OL stage.

Fire hazards (for both sodium and conventional fires) are highly dependent on the final design configuration and plant construction. The coordination of construction activities with the development of the fire protection strategy is within the purview of the applicant, including choices to manage risks during the construction process. These activities may be included within the scope for construction oversight, and the Staff will maintain awareness of these activities through implementation of the construction oversight process. The Staff has also encouraged the applicant to engage in pre-application interactions regarding fire protection to ensure the construction activities align with the fire protection strategy based on the final design, to improve regulatory clarity and efficiency at the OL stage. The applicant has noted planned interactions in these areas in their most recent regulatory engagement plan, including a planned white paper on its approach to fire protection for the Natrium design (ML25273A340).

11 American Nuclear Society, A new ANSI/ANS standard for liquid metal fire protection published (Sept.

8, 2025), https://www.ans.org/news/2025-09-08/article-7349/a-new-ansians-standard-for-liquid-metal-fire-protection-published/

12 When referencing ANS-54.8 without a date, the Staff is discussing concepts common to both the 1988 and 2025 versions of the standard.

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7. License conditions 4.G, 4.H, and 4.I require reports to be submitted for research and development (R&D) activities. How will the Staff use these reports prior to the OL application review? Are there other mechanisms to obtain the information? Is it possible to modify the license conditions to allow for more flexibility or to reduce the burden of the submitted reports?

Staff Response: The Staff identified that the topics specified in these permit conditions cover areas that represent sufficient nuclear safety and regulatory risk to warrant focused engagement during the construction phase to provide reasonable assurance that such safety questions will be satisfactorily resolved in the final design. The Staffs approach to address considerations like this through permit conditions allowed the Staff to assess the preliminary design in a manner that maximized flexibility to the permit holder. The Staff will use the reports on research and development (R&D) activities to understand the results of these activities and the resolution of the associated safety questions, in accordance with 10 CFR 50.35(a). The reports will provide the Staff the opportunity to offer feedback to USO on the progress of the R&D activities, which will support Staffs future review of the final design, and contribute to mitigating future licensing uncertainties at the OL stage.

For example, USO plans to preclude the possibility of significant sodium-salt interactions through the design of the sodium-salt heat exchanger (SHX). The R&D activities will help to develop and refine the applicants approach and could result in changes to the SHX design. The R&D reports will provide assurance that design changes to the SHX will resolve related safety questions and may inform the construction oversight activities. For instance, it may be prudent for the Staff to prioritize construction oversight on the SHX fabrication process because certain safety issues may only be identifiable during this phase and not be easily detected later during operation.

The Staff discussed these permit conditions with the applicant and confirmed that the information was consistent with USOs R&D plans. Additionally, consistent with the principles of good regulation, the Staff confirmed with USO that the information requested by the permit conditions and their tie to regulatory requirements was clear, and that the timing and content of the requested reports would not impose an undue burden. Other mechanisms that could be used to obtain the information include the voluntary pre-application engagement process (e.g.,

applicant submittal of white papers or topical reports) or oversight (e.g., NRC inspectors performing oversight of the quality assurance process for construction could obtain access to some of this information). However, while it may be possible to modify the license conditions to allow for more flexibility or reduce the reporting burden, the Staff determined that the proposed approach appropriately optimizes efficiency, clarity, and flexibility in a manner consistent with 10 CFR 50.35.

8. Pursuant to 10 C.F.R. § 50.35(b), the Staff has included license conditions 4.G and 4.H requiring annual reporting of R&D activities for the Intermediate Heat Transport System Sodium-Salt Heat Exchanger Interaction (SHX) and the Reliability and Integrity Management (RIM) program. It appears that due to the importance of these R&D activities, clarity on the outcome of the R&D is desired to support the Staffs determination on the readiness and acceptance of the OL application. However, given the stated purpose of the relevant license conditions and the NRCs regulatory footprint after issuance of the CP, it is unclear whether

11 annual tracking of the R&D provides additional benefit commensurate with the regulatory burden compared to obtaining a summary report after the completion of the R&D program that identifies any major changes compared to the CP.

Please consider the example below for an alternative to the current license condition 4.G:

Prior to the completion of construction and after the completion of research and development (R&D) activities associated with the sodium-salt heat exchanger design and sodium-salt reactions, USO shall submit a summary report to the NRC that provides an overview of the completed R&D program, including, (i) key activities to characterize the sodium-salt reaction, mature and develop the SHX design, develop appropriate design features and controls needed to prevent and mitigate sodium-salt reactions, and materials testing, to improve understanding of the effects of high temperature and exposure to the sodium and salt environment on SHX materials, including weld metals and diffusion bonded material, (ii) the key results from the activities identified in item (i), and (iii) any differences or changes in behavior and design due to the R&D activities compared to the information presented in the PSAR, Revision 1.

a. Would the alternative license condition achieve the Staffs intent? If not, please explain why. If the alternative license condition is acceptable, please submit a revised License Condition 4.G for the Commissions consideration.

Staff Response: No, receipt of an update after the completion of R&D activities as proposed in the alternative condition would not achieve the Staffs intent for License Condition 4.G. As discussed in greater detail in response to question 7, the reports will provide the Staff the opportunity to offer timely feedback to USO on the progress of the R&D activities in resolving important safety questions, and inform the construction oversight process. Receiving a summary report after the completion of research and development (R&D) activities would not achieve these goals, because the information in the Staffs proposed periodic reports would support ongoing activities.

b. Would a similar alternative approach achieve the Staffs intent for License Condition 4.H? If not, please explain why. If so, please submit a revised License Condition 4.H for the Commissions consideration.

Staff Response: No, a similar alternative license condition to receive an update after the completion of R&D activities would not achieve the Staffs intent for License Condition 4.H for the reasons provided in response to subpart a.

9. Pursuant to 10 C.F.R. § 51.105(a)(1), the Commission must determine in this proceeding whether the requirements of NEPA section 102(2)(C) have been met.

Pursuant to section 51.105(a)(4), the Commission must also determine whether the NEPA review conducted by the Staff has been adequate.

NEPA section 102(2)(C)(v), as amended by the Fiscal Responsibility Act of 2023

12 (FRA), requires an agencys environmental review to consider any irreversible and irretrievable commitments of Federal resources which would be involved in the proposed agency action should it be implemented. In February 2025, CEQ stated that until agencies complete appropriate revisions to their NEPA-implementing procedures, agencies should apply their current NEPA implementing procedures with any adjustments needed to be consistent with the NEPA statute as revised by the FRA. CEQ reiterated this guidance in a September 2025 memorandum.

In SECY-24-0046, the Staff affirmed that it planned to limit the discussion of irreversible and irretrievable commitments of resources in EISs to Federal resources.

However, it is not clear that the EIS or Staffs Information Paper discuss this specific topic. For example, FEIS section 6.3.3 only discusses irreversible and irretrievable commitments of resources; it does not identify which, if any, of these are Federal resources. Likewise, the Staffs Information Paper simply states that the FEIS addresses any irreversible and irretrievable commitments of resources that would be involved in the proposed action.

a. Please clarify:

1) Whether the Staff has identified any irreversible and irretrievable commitments of Federal resources that would be involved in implementing the proposed agency action.

2) If any such Federal resources have been identified, how they have been addressed in the FEIS.

b. If necessary, consistent with NEPA section 102(2)(C)(v), as amended by the FRA, the Staff should supplement the FEIS by responding to this question with a summary of any irreversible and irretrievable commitments of Federal resources which would be involved in the proposed agency action.

For example, such resources may include those expended by the Staff in its review of the project or the expenditure of funds from the U.S.

Department of Energy to support the proposed action.

Staff Response: The National Environmental Policy Act of 1969, as amended, requires that major Federal actions significantly affecting the quality of the human environment include a detailed statement on, in part, any irreversible and irretrievable commitments of Federal resources that would be involved in the proposed agency action should it be implemented. In its final environmental impact statement (FEIS) for the construction permit application for Kemmerer Unit 1, the Staff discusses irreversible and irretrievable commitments of resources, without distinction between Federal and non-Federal resources. In Section 6.3.3 of the FEIS, the Staff states that an irreversible commitment of resources occurs when potential impacts have the possibility to limit future options for a resource and that an irretrievable commitment of resources is the lost production or use of a resource that would cause the resource to be unavailable for use by future generations. Based on these definitions, the Staff then summarizes the irreversible and irretrievable commitments of resources that have been noted in the FEIS.

13 (a) In response to the Commissions question, the Staff reassessed the issue of irreversible and irretrievable commitments of resources. The Staff confirmed, based on its definitions of irreversible and irretrievable, that Section 6.3.3 of the FEIS accurately summarizes the irreversible and irretrievable commitments of resources that would be involved in the issuance of a construction permit for Kemmerer Unit 1. And with respect specifically to Federal resources, the Staff confirmed, based on its definitions of irreversible and irretrievable, that there are no irreversible and irretrievable commitments of Federal resources that are omitted from the FEIS.

(b) Because the Staff determined that no irreversible and irretrievable commitments of Federal resources are omitted from the FEIS, the Staff believes that it is not necessary to supplement the FEIS.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the Matter of US SFR OWNER, LLC (Kemmerer Power Station, Unit 1)

Docket No. 50-613-CP CERTIFICATE OF SERVICE Pursuant to 10 C.F.R. § 2.305, I hereby certify that copies of the foregoing, NRC STAFF RESPONSES TO COMMISSION HEARING QUESTIONS, dated January 26, 2026, have been served upon the Electronic Information Exchange, the NRCs E-Filing System, in the above-captioned proceeding, this 26th day of January 2026.

/Signed (electronically) by/

Julie G. Ezell Counsel for NRC Staff Mail Stop: O15-B04 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Telephone: (301) 287-9157 E-mail: julie.ezell@nrc.gov