ML26021A186

Us Sfr Owner, LLC Responses to Commission Hearing Questions

ML26021A186
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	01/21/2026
From:	Lighty R, Polonsky A, George Wilson Morgan, Morgan, Lewis & Bockius, LLP, US SFR Owner
To:	NRC/OCM
SECY RAS
References
RAS 57580, 50-613-CP
Download: ML26021A186 (0)

Text

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 January 21, 2026 US SFR OWNER, LLC RESPONSES TO COMMISSION HEARING QUESTIONS US SFR Owner, LLC (USO) provides the following responses to the questions in the Commissions January 12, 2026, Order (Transmitting Hearing Questions) regarding the mandatory hearing for the Construction Permit Application for the Kemmerer Power Station, Unit 1. USO responses are limited to those questions directed to it.

Question 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 Staff's guidance on PRA acceptability in Regulatory Guide (RG) 1.247, the Staff's 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.

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

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?

USO Response to Question 3

a. The NEI 18-04 methodology defines an iterative process and is therefore well suited to incorporate changes throughout project design and construction phases.

The design control program implemented in accordance with 10 CFR 50 Appendix B, Criterion III and the NRC approved TerraPower Quality Assurance Program Description (ADAMS Accession No. ML23213A199) ensures that the conditions of the final as-built plant are consistent with the associated analytical calculations. The design control program implements the selection and evaluation of Licensing Basis Events, safety classification of structures, systems, and components (SSCs) and associated risk-informed special treatments, and determination of defense-in-depth adequacy performed in accordance with NEI 18-04, Revision 1, as endorsed by Regulatory Guide 1.233.

As noted in the question, iterations of the NEI 18-04 methodology may result in changes to the safety classification of SSCs and the associated required special treatments. The NEI 18-04 methodology also includes iterative evaluation of defense-in-depth adequacy by the integrated decision-making process (IDP) and integrated decision-making process panel (IDPP). The IDP and IDPP may determine that additional requirements on SSCs, which could include elevation of safety classification, must be implemented for defense-in-depth adequacy.

If the safety classification of an SSC is changed, the appropriate revised special treatments for the SSC will be selected in accordance with the NEI 18-04 methodology.

If an SSC safety classification is elevated to a safety classification with higher safety significance (e.g., non-safety-related to safety-related), methods such as commercial grade dedication, additional inspection or testing, modification, or replacement may be considered to confirm the SSC conforms to the revised safety classification and associated special treatments. The applicant made conservative assumptions at the construction permit (CP) stage intended to reduce the likelihood that a non-safety-related SSC will become safety-related at the operating license (OL) stage. The final safety analysis report approved by the Commission will reflect the design and analyses of the as-built plant, including appropriate incorporation of changes made during design finalization and construction.

3

b. In accordance with 10 CFR 50.35(b), changes to 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 may be subject to an amendment of the construction permit using 10 CFR 50.90. As the Commission has recognized, conditions for which an amendment to a construction permit is required for such changes are not defined in the Atomic Energy Act or NRC regulations.1 In this context, proposed changes historically have been evaluated on an ad hoc basis and the need for construction permit amendments turns on the significance of the change.2 However, as an example, an amendment may be prudent if an SSC in the preliminary safety analysis report is changed to a safety classification of lower safety significance, or if there is a consequential revision to the principal design criteria.

As described in 10 CFR 50.35(c), the NRC will not issue a license authorizing operation of any facility until (1) the applicant submits, as part of an OL application, its final safety analysis report and (2) the Commission finds that the final design provides reasonable assurance that operation of the facility in accordance with the requirements of the license and NRC regulations will not endanger public health and safety. The final safety analysis report approved by the Commission will reflect the design and analyses of the as-built plant, including appropriate incorporation of changes made during design finalization and construction.

In addition, the Post-Construction Inspection, Test and Analysis Program (PITAP) described in Chapter 12 of the preliminary safety analysis report will provide reasonable assurance that the plant is constructed as designed and will operate within the bounds of the approved final safety analysis report. NRC oversight of quality and adherence of the as-built plant to the approved licensing basis is achieved through the NRC advanced reactor construction oversight program (ARCOP), which evaluates the quality of the fabrication, manufacturing, and construction of advanced reactor projects using a continuous assessment process.

1 N. Ind. Pub. Serv. Co. (Bailly Generating Station, Nuclear-1), CLI-79-11, 10 NRC 733, 735 (1979). See also Domestic Licensing of Production and Utilization Facilities; Design and Other Changes in Nuclear Power Plant Facilities After Issuance of Construction Permit, 45 Fed. Reg. 81,602 (Dec. 11, 1980) (advanced notice of proposed rulemaking (ANPRM) that would have clarified when changes require a construction permit amendment); but see also Regulatory Agenda, 51 Fed. Reg. 14,880, 14,912 (Apr. 21, 1986) (withdrawing that ANPRM because, among other reasons, all construction permit holders are now operating license applicants, rendering the issue moot at that time).

2 Bailly, CLI-79-11, 10 NRC at 735 (Depending on the degree of significance, a proposed change may receive detailed staff review, but more commonly, detailed review is deferred to the operating license review stage.

Although a sufficiently major change could warrant a construction permit amendment, a review of 88 extant construction permits indicated that none had been amended for a design change.). See also Bailly, NRC Staff Response to Commission Questions of December 11, 1978 at 10-12 (Jan. 10, 1979) (ML19256A923)

(describing other informal ways the NRC Staff considers such changes).

4 Question 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 Applicant's 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?

USO Response to Question 6 Yes, USO has assessed the availability of design standards for detection, suppression, and support systems that would be applicable to Kemmerer Power Station. USO used ANSI/ANS 54.8-2025, Liquid Metal Fire Protection in Liquid Metal Reactor Plants, as guidance for the design of SSCs needed to prevent and mitigate sodium fires. In addition, USO is using IEEE standards for the design of electrical components (e.g., detectors) and ASME standards for active and passive mechanical components. Despite most guidance being developed for LWR plants (i.e., RG 1.189, Fire Protection for Nuclear Power Plants, and NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities), the concepts contained therein are generally applicable to non-LWR designs. Based on the design standards identified, there exists sufficient guidance for the design of detection, suppression, and support systems used at Kemmerer Power Station Unit 1.

USOs response to question 3 describes the process for implementing changes to design and safety classification of SSCs identified during construction. Preliminary analyses for the fire protection strategy were used to inform the design and plant layout as described in the preliminary safety analysis report. As construction progresses and the design is finalized, any design changes identified will follow the design control process in accordance with the NRC approved TerraPower Quality Assurance Program Description (ADAMS Accession No. ML23213A199) and applicable procedures. The design control program implemented in accordance with 10 CFR 50 Appendix B, Criterion III ensures that the conditions of the final as-built plant are consistent with the associated analytical calculations.

5 Question 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 Staff's determination on the readiness and acceptance of the OL application.

However, given the stated purpose of the relevant license conditions and the NRC's regulatory footprint after issuance of the CP, it is unclear whether 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.

For the Applicant:

a.

Does the alternative license condition present any obstacles compared to the current License Condition 4.G?

b.

Does an alternative license condition with a similar approach present any obstacles compared to the current license condition 4.H?

USO Response to Question 8

a.

No, the alternative license condition does not present any obstacles compared to the current License Condition 4.G.

b.

No, the alternative license condition does not present any obstacles compared to the current License Condition 4.H.

6 Respectfully submitted, Executed in Accord with 10 C.F.R. § 2.304(d)

GEORGE WILSON Senior Vice President, Regulatory Affairs US SFR OWNER, LLC 15800 Northup Way Bellevue, WA 98008 Phone: 425.324.2888 Email: gwilson@terrapower.com Signed (electronically) by Ryan K. Lighty RYAN K. LIGHTY, ESQ.

MORGAN, LEWIS & BOCKIUS LLP 1111 Pennsylvania Avenue, N.W.

Washington, D.C. 20004 Phone: 202.739.5274 Email: Ryan.Lighty@morganlewis.com Executed in Accord with 10 C.F.R. § 2.304(d)

ALEX S. POLONSKY, ESQ.

MORGAN, LEWIS & BOCKIUS LLP 1111 Pennsylvania Avenue, N.W.

Washington, D.C. 20004 Phone: 202.739.5830 Email: Alex.Polonsky@morganlewis.com Counsel for US SFR Owner, LLC

DB1/ 165687729.3 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 January 21, 2026 CERTIFICATE OF SERVICE I hereby certify that, on this date, a copy of the foregoing US SFR OWNER, LLC RESPONSES TO COMMISSION HEARING QUESTIONS was served upon the Electronic Information Exchange (the NRCs E-Filing System) in the above-captioned docket.

Signed (electronically) by Ryan K. Lighty RYAN K. LIGHTY, ESQ.

MORGAN, LEWIS & BOCKIUS LLP 1111 Pennsylvania Avenue, N.W.

Washington, D.C. 20004 Phone: 202.739.5274 Email: Ryan.Lighty@morganlewis.com Counsel for US SFR Owner, LLC