ML23257A258

Terrapower, LLC – Final Safety Evaluation of Topical Report NATD-LIC-RPRT-0001, Regulatory Management of Natrium Nuclear Island and Energy Island Design Interfaces, Rev. 0

ML23257A258
Person / Time
Site:	99902100
Issue date:	10/04/2023
From:	Jessup W NRC/NRR/DANU/UAL1
To:	TerraPower
Sutton M
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ML23257A260	List: ML23257A258 ML23257A259
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Enclosure TERRAPOWER, LLC - FINAL SAFETY EVALUATION OF TOPICAL REPORT NATD-LIC-RPRT-0001, REGULATORY MANAGEMENT OF NATRIUM NUCLEAR ISLAND AND ENERGY ISLAND DESIGN INTERFACES, REVISION 0 (EPID: L-2022-TOP-0045)

SPONSOR AND SUBMITTAL INFORMATION Sponsor:

TerraPower, LLC Sponsor Address:

15800 Northup Way, Bellevue, WA 98008 Project No.:

99902100 Submittal Date: October 4, 2022 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML22277A824 [1]

Brief Description of the Topical Report: The topical report (TR) describes an evaluation of the NRC regulations pertaining to the design interface of the Nuclear Island [NI] and Energy Island

[EI] systems for the Natrium' design and includes a supporting summary of the Natrium reactor plant design, interfaces, safety features, and basic plant transient analysis, and TerraPowers process for classifying structures, systems, and components (SSCs). TerraPower stated in the TR that [t]he independence of operation between the systems contained within the NI and the plant systems composing the EI is a key aspect of the Natrium design philosophy.

To support the implementation of this philosophy, [t]he NI boundary conditions have been intentionally designed so the interrelationship with the EI does not impact the NI safety case.

In the TR, TerraPower evaluated four regulations in Title 10 of the Code of Federal Regulations (10 CFR) as they pertain to the interfaces between the NI and EI systems and the design of the Natrium reactor. For two of the regulations, TerraPower intends to request exemptions and provided brief discussions of the planned technical and regulatory rationales for these exemptions. However, the exemptions themselves were not requested in the TR.

TerraPower requested NRC staff review and approval of the subject TR to serve as a means, via reference, for Natrium reactor licensees to utilize the regulatory evaluation.

By email dated November 17, 2022 [2], (ML22319A153), the NRC staff informed TerraPower that the TR provided sufficient information for the NRC staff to conduct a detailed technical review [3].

On January 5, 2023, the NRC staff submitted an audit plan to TerraPower [4], (ML22353A642),

and subsequently conducted an audit of materials related to the TR from January 23, 2023, to March 10, 2023. The audit summary was issued on June 16, 2023 [5], (ML23167A478).

REGULATORY EVALUATION The NRC staff reviewed whether the regulatory analysis provided in the TR was consistent with the relevant statutes, regulations, guidance, and Commission policy, and supportable by the Natrium design in its current state as discussed in the TR.

The statute with requirements relevant to the operator and operator license issues discussed in the TR is the Atomic Energy Act of 1954, as amended (AEA), which states, in part, the following:

Section 11, Definitions, Paragraph r.: The term operator means any individual who manipulates the controls of a utilization of production facility.

Section 107, Operators Licenses: The Commission shall prescribe uniform conditions for licensing individuals as operators of any of the various classes of production and utilization facilities licensed in this Act.

The regulations evaluated by TerraPower in the TR and assessed by the NRC staff include:

10 CFR 50.10, License required; limited work authorization, which in the pertinent part discusses the types of activities that constitute construction and may not be conducted without a construction permit (CP) or limited work authorization; 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (also known as the Maintenance Rule), which provides requirements for monitoring nuclear power plant maintenance activities, including requirements regarding the scope of monitoring programs; 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, which establishes quality assurance requirements relevant to nuclear power plants; and 10 CFR Part 55, Operators Licenses, which establishes requirements for operators licenses.

The NRC staff also considered the following regulations:

10 CFR 50.2, Definitions, which provides definitions for:

o Safety-related structures, systems and components o Controls for nuclear reactors 10 CFR 50.12, Specific exemptions, which provides requirements associated with exemptions from the regulations 10 CFR 50.54, Conditions of licenses, specifically paragraphs (i) and (j). Paragraph 50.54(i) allows only licensed operators or senior operators to manipulate the controls of a reactor. Paragraph 50.54(j) provides requirements for the manipulation of [a]pparatus and mechanisms other than controls, the operation of which may affect the reactivity or power level of a reactor[.]

The guidance considered by the NRC staff includes:

Regulatory Guide (RG) 1.206, Revision 1, Applications for Nuclear Power Plants

[6]. Though this document primarily relates to applications submitted under the processes in 10 CFR Part 52, Section C.2.18 provides guidance on limited work authorizations (LWAs) and the types of activities constituting or not constituting construction under 10 CFR 50.10, with examples.

RG 1.160, Revision 4, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants [7], which provides guidance on implementation of the Maintenance Rule (10 CFR 50.65).

RG 1.233, Revision 0, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors [8], which provides guidance on using a technology-inclusive, risk-informed, and performance-based methodology to inform the licensing basis and content of applications for non-light-water reactors (non-LWRs). This RG endorses NEI 18-04, Revision 1, Risk-Informed Performance-Based Guidance for Non-Light Water Reactor Licensing Basis Development [9], which provides an approach to selecting licensing basis events (LBEs), classifying structures, systems and components (SSCs), and assessing defense-in-depth (DID).

TECHNICAL EVALUATION 1.0 NATRIUM PLANT DESIGN AND TRANSIENTS The plants design is fundamental to the transients that may occur at the facility, the plant response to those transients, and the safety classification of the SSCs that prevent and mitigate those transients. These are significant considerations in evaluating TerraPowers regulatory analyses seeking to demonstrate the independence of the Natrium NI and EI. Sections 1.1 through 1.8 of this safety evaluation (SE) provide context for the regulatory analyses assessed in Section 3.0 of this SE. The NRC staff considered the plant design, as discussed in Sections 3, 4, and 5 of the TR, and the transient response of the reactor, as discussed in Section 7 of the TR, and highlighted those design features that are significant to the independence of the Natrium NI and EI. The NRC staff has placed a condition on the use of this TR, as outlined in the Limitations and Conditions section of this SE, to ensure that these key design features are appropriately addressed in future licensing applications or correspondence referencing this TR.

1.1 Fuel and core As discussed in Sections 2, 3, 4, and 5 of the TR, the proposed Natrium reactor is a metal-fueled, pool-type sodium fast reactor (SFR). The fuel and core are located on the NI. The proposed initial Natrium core uses fuel composed of metallic uranium alloyed with 10 weight percent zirconium (U-10Zr). The fuel is formed in slugs and inserted into fuel rods made of HT9, a martensitic stainless steel alloy. The fuel is bonded to the fuel rod cladding with sodium, which fills the gap between the fuel and cladding and improves thermal conductivity from the fuel out to the coolant. These fuel rods are arranged into assemblies and inserted into the core. Other types of assemblies in the core include control assemblies, reflector assemblies, and shield assemblies. The fuel operates in the fast neutron spectrum. The core is cooled by liquid metallic sodium, which circulates between cold and hot pools within the vessel as discussed in more detail in Section 1.2 below.

The NRC staff determined that the designs of the fuel and core, as described in the TR, are key aspects of the plant for NI-EI independence. The metallic fuel has a high thermal conductivity and thus peak fuel temperatures are expected to be close to coolant temperatures and below the coolant boiling point. The NRC staff expects the fuel system to have substantial margin to safety limits, even during transients, and beneficial reactivity feedback effects; these features are necessary to avoid damage to the fuel system because of transients. Additionally, the fuel system has beneficial characteristics for retention of radionuclides in the event of an accident. All these aspects of the fuel systems behavior are necessary for NI-EI independence because they improve safety margins and prolong transients, allowing NI systems to respond to mitigate them without reliance on the EI.

1.2 Primary heat transport system and reactor vessel Section 5.1 of the TR describes the reactor vessel and the Natrium primary heat transport (PHT) system. The reactor vessel and PHT are located on the NI. In the PHT, liquid sodium is circulated around the reactor vessel by mechanical primary sodium pumps (PSPs), which take suction from the cold pool and discharge into the core inlet. From there, the sodium is heated by the core and flows into the hot pool. Hot sodium then flows downward through the shell side of the intermediate heat exchangers (IHXs) where its energy transfers into the intermediate heat transport (IHT) system. The cooled sodium then enters the cold pool and completes the flow circuit. An illustration of the PHT system is in Figure 1 of the TR.

The reactor vessel is surrounded by a guard vessel, which is designed to contain sodium leaks in the event of a breach in the reactor vessel. Between the reactor vessel and the guard vessel is a small gap filled with inert gas.

The NRC staff determined that the design of the primary heat transport system, as described in the TR, is a key aspect of the plant for NI-EI independence because (a) it contains the coolant and provides the geometry necessary for coolant to flow past the core, (b) it provides the primary pathway for heat to be removed from the fuel by the reactor air cooling (RAC) and intermediate air cooling (IAC) systems (discussed below in Section 1.6), and (c) it provides thermal inertia that insulates the reactor core from upsets on the EI.

1.3 Intermediate heat transport system Section 5.1 of the TR also discusses the IHT system. The IHT is located on the NI. In the IHT system, liquid sodium is circulated by mechanical intermediate sodium pumps (ISPs) from the IHX tube side to the tube side of the sodium-salt heat exchangers (SHXs). There are two IHT loops, each of which is connected to one IHX and two SHXs. Between the IHX and the SHX, the sodium passes through sodium-air heat exchangers (AHXs), which are a part of the IAC system. The IHT also contains expansion and drain tanks to handle the sodium. The IHT system can be seen in TR Figure 2.

The NRC staff determined that the design of the IHT system, as described in the TR, is a key aspect of the plant for NI-EI independence because (a) it provides a pathway for heat to be removed from the fuel by the IAC (discussed below in Section 1.6) and (b) it provides additional thermal inertia that reduces the effects on the reactor core resulting from transients on the EI.

1.4 Thermal salt system and boundary between NI and EI Section 5.1 of the TR states that the four SHXs allow heat to be transferred from the IHT system, which contains sodium, to the thermal salt system (TSS), which contains a molten salt similar to that used in concentrated solar systems. In the TSS, variable speed mechanical pumps take suction from the cold salt tank and pump molten salt through the shell side of the SHXs, where it is heated by the IHT, and into a hot salt storage tank. In the SHXs, salt pressure is higher than sodium pressure, so leakage across the SHX tubes would be salt from the TSS into the sodium in the IHT. The TSS is on both the NI and EI. As discussed in Sections 5.1 and 7 of the TR, the boundary between the EI and NI is provided by drain isolation valves on the inlet and outlet of the salt side of the SHXs. The SHXs, a small amount of salt piping, and the salt isolation valves are therefore considered to be part of the NI, while the salt tanks and the salt piping on the EI-side of the isolation valves are considered to be part of the EI. As discussed in Section 7 of the TR, the drain isolation valves close in response to transients, isolating the NI from the EI.

The NRC staff determined that the TSS, as described in the TR, is a key aspect of the plant for NI-EI independence because it plays a key role in allowing the NI and EI to operate independently. The NI portion of the TSS, which includes the SHXs, the salt drain isolation valves, and the piping in between, may be relied on for safety; this portion of the TSS allows the NI to be isolated from the EI to mitigate effects of failures in the EI portion of the salt system on the SHXs.

Also, the large volume in the thermal salt tanks in the EI portion of the TSS reduces the effects on the NI resulting from transients in the steam generation, power conversion, or other EI systems (e.g., power output changes, turbine trips). However, the NRC staff notes that while the TSS plays a key role in the independent operation of the NI and EI, the EI portion of the TSS is not relied on for the safety of the reactor. This is because the NI can effectively respond to transients regardless of the condition of the EI portion of the TSS, as discussed in Section 7 of the TR and Section 1.8 of this SE.

1.5 EI systems EI systems are discussed in Section 5.2 of the TR. These include the TSS, the steam generation system, the condensate and feedwater system, the turbine and generator and associated systems, and the heat rejection system. A diagram of some of these systems is provided in Figure 5 of the TR.

Salt is pumped from the hot and cold salt tanks (mixed to maintain desired temperatures) through an economizer, evaporator, superheater, and reheater, to generate steam to run the turbine. Salt that has deposited energy in the steam generation system is returned to the cold salt tank so it may be reheated by the SHXs.

Steam used to drive the turbine goes through a condenser, where it is subcooled, and passes through feedwater heaters and a deaerator before being pumped back to the steam generation system. The turbine itself is planned to be a commercially available steam turbine. Steam can bypass the turbine to the condenser. The condenser is cooled by a circulating water system that rejects heat to the atmosphere via a mechanical forced draft cooling tower.

The NRC staff determined that EI systems other than the TSS are not significant to NI-EI independence, because their effects on the core are reduced by the thermal inertia provided by the TSS, IHT system, and PHT system.

1.6 Reactor air cooling and intermediate air cooling systems The reactor air cooling (RAC) and intermediate air cooling (IAC) systems are discussed in Section 5.1 of the TR and illustrated in TR Figure 3. These systems provide residual heat removal capability for the Natrium reactor.

The RAC is a natural draft system that utilizes ducts to guide outside air past the guard vessel.

Heat from the core is transported to the reactor vessel by a mix of convection (either forced or natural depending on the state of the plant) and conduction, then from the reactor vessel to the guard vessel primarily by radiation

1. The exterior surface of the guard vessel is cooled by a mixture of convection from the RAC air flow and radiation to surrounding structures (primarily the collector cylinder), which are also cooled by the RAC air flow. The RAC is a fully passive system with no dampers and is therefore always in operation.

The IAC, as mentioned previously in Section 1.3 of this SE, consists of an AHX on the IHT system along with blowers and dampers on the air side of the AHX. In active mode, the IAC provides controlled, forced flow of air across the AHX that allows heat to be transferred from the IHT to the atmosphere. In passive mode, the dampers are designed to fail open so that natural draft flow of outside air can remove heat from the IHT without active systems, as a supplement to the RAC.

As discussed in Section 4.2 of the TR, the IAC (in forced flow mode) is intended to serve as the normal shutdown cooling system, while the RAC is designed to be able to remove all decay heat from the reactor using purely passive means. The NRC staff notes that because radiation heat transfer is a key contributor to RAC performance, the reactor may need to heat up for the RAC to remove heat effectively. The NRC staff audited preliminary analyses performed by TerraPower that included these systems and verified that they appear to be capable of performing the necessary heat removal function, based on the current design. The NRC staff stresses, however, that the analyses audited were preliminary and are subject to confirmation as design details are developed and finalized.

The NRC staff determined that the RAC and IAC, as described in the TR, are a key aspect for NI-EI independence because they provide the capability to remove all decay heat using systems located solely on the NI.

1.7 Reactor control and protection systems Reactivity control is accomplished via control rods and inherent feedback mechanisms, as discussed in part in Section 1.1 of this SE. Shutdown is accomplished via control rods. As stated in Section 4.1 of the TR, all sensors monitoring reactor trip parameters, except for some seismic sensors, are located on the NI. A non-safety related anticipatory power runback feature is also included in the design which inserts the control rods automatically to reduce power and attempt to avoid a scram, providing additional time and margin to respond to transients.

However, as discussed in Section 4.1 of the TR, if a reactor runback or scram fails, the core is designed such that many anticipated transients can be accommodated without scram.

The NRC staff determined that the reactor control and protection systems, as described in the TR, are a key aspect for NI-EI independence because they are designed to be able to shut down the reactor without relying on SSCs, including sensors, located on the EI.

1.8 Natrium transients TerraPower stated in Section 7 of the TR that no EI SSCs are required to (1) respond to mitigate any events impacting the NI, (2) support safety-related (SR) SSCs, or (3) ensure DID adequacy.

The response to transients impacting the NI can be accomplished solely with 1 The inert gas in the gap between the reactor vessel and guard vessel may also provide a minor contribution to heat transfer through conduction or natural convection. However, radiation is expected to be the dominant means of heat transfer across the gap.

equipment located on the NI. Section 7 of the TR also describes the response to transients initiated on the EI.

Specifically, from the NI perspective, all transients on the EI can be grouped into either increases in heat removal by the salt system (increased salt flow, low salt temperature) or decreases in heat removal by the salt system (decreased salt flow, high salt temperature, low salt pressure). These manifest as a decrease or increase, respectively, in the IHT cold leg temperatures. With no action, this eventually causes an increase or decrease, respectively, in the core inlet temperature.

TerraPower stated the thermal inertia of the IHT and PHT is such that changes in salt conditions can be adequately responded to using only signals monitored within the NI; this was verified by the NRC staff in the audit. As discussed in Section 7.1 of the TR, the plant isolates the NI from the EI after a scram or runback by closing the SHX drain isolation valves, so there is no prolonged effect on the NI from any EI systems. As discussed in Section 1.6 of this SE, TerraPowers preliminary analyses further show that removal of all decay heat can be accomplished solely with NI systems (RAC and IAC).

The NRC staff therefore determined that, based on the summary of Natrium design and analyses presented in the TR, the NI has the capacity to effectively respond safely to transients, regardless of whether they are initiated on the NI or EI, using only NI systems. This plays a key role in the safety classification of SSCs on the EI, which is discussed in further detail in Section 2.0, below.

2.0 NATRIUM SAFETY CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS As discussed in Section 1 of the TR, the Natrium reactor licensing approach follows NEI 18-04 [9], which was endorsed by the NRC staff in RG 1.233 [8]. NEI 18-04 provides a risk-informed and performance-based process for determining the safety classification of SSCs at nuclear reactor facilities. The safety classification process in NEI 18-04 is highly integrated with the rest of NEI 18-04, which also provides processes for selecting LBEs to be included in the plant safety analysis2 and ensuring adequate DID. These other processes will only be discussed here to the extent that they have a direct nexus to the SSC classification.

The safety classification of an SSC is used to determine the standards to which it is designed, fabricated, and maintained. Additionally, certain regulatory requirements apply to SR SSCs and other SSCs that are deemed important to safety; some of these requirements will be discussed later in this SE.

Figure 6 of the TR provides a flow chart with the NEI 18-04 SSC classification process.

Additionally, a high-level overview of the ultimate safety-classification categories is provided in Section 4 of NEI 18-04. The categories are:

2 The LBE selection and evaluation process described in NEI 18-04 identifies the transients that must be analyzed for the facility based on the plant design and a probabilistic risk assessment (PRA). Transients are assigned to LBE categories based on frequency. Anticipated operational occurrences are events with frequency greater than 10-2 per year; design basis events (DBEs) are those with frequency between 10-4 and 10-2 per year; and beyond design basis events (BDBEs) are those with frequency between 5*10-7 and 10-4 per year. The LBEs are evaluated against a frequency-consequence curve derived from regulatory requirements for dose. Design-basis accidents (DBAs) are derived from the DBEs by analyzing them to ensure that they meet the 10 CFR 50.34 dose limits, applying conservative assumptions and only crediting equipment selected to be safety-related in the analysis.

SR, which includes:

o SSCs selected3 to perform safety functions required to mitigate the consequences of DBEs to within the frequency-consequence target, o SSCs selected to mitigate DBAs (which rely solely on SR SSCs) to meet the dose limits of 10 CFR 50.34 using conservative assumptions, and o SSCs selected to prevent the frequency of high-consequence BDBEs4 from increasing into the DBE region and beyond the frequency-consequence target curve Non-safety related with special treatment (NSRST), which includes:

o Non-safety-related SSCs relied on to perform risk-significant functions; that is, those functions that prevent or mitigate any LBE from exceeding the frequency-consequence target or make significant contributions to the cumulative risk metrics (discussed in more detail in Section 3.3.6 of NEI 18-04), and o Non-safety-related SSCs relied on for DID adequacy Non-safety related with no special treatment (NST), which includes all other SSCs.

In addition to these definitions provided in NEI 18-04, the NRC staff clarified in RG 1.233 that the NRC expects that SSCs that provide essential support (including required human actions) for SR or NSRST SSCs will be classified in a manner consistent with the higher-level function, even if the supporting SSC is not explicitly modeled in the PRA.

TerraPower determined that all SSCs on the EI would be NST according to the NEI 18-04 process because they are not needed to meet the required safety functions, do not provide risk-significant functions, and are not needed for DID. As described in Section 1.0 of this SE, the NRC staff reviewed TR summaries of proposed Natrium design features and basic plant transient responses as well as associated supporting analysis during a regulatory audit. With respect to DID considerations, TerraPower indicated in TR Sections 4.1 through 4.3 that there are multiple NI SSCs capable of accomplishing the plant safety functions and providing adequate DID without involving the EI. Based on the NRC staffs review of this information, the NRC staff determined that a future applicant could be able to justify characterizing EI systems as NST for a Natrium design with these specified high-level features. The need for future applicants and licensees referencing the TR to demonstrate that EI SSCs are appropriately classified as NST is listed as one of the limitations and conditions in the Limitations and Conditions section of this SE.

Additionally, the NRC staff noted in RG 1.233 that the definition of safety-related SSCs in NEI 18-04 differs from that provided in 10 CFR 50.2. While RG 1.233 determined that the process provided in NEI 18-04 was acceptable, it also noted that [a]pplicants referencing this RG are 3 The term selected is used because multiple, overlapping sets of SSCs may be able to carry out the necessary safety functions. Thus, one set of SSCs that is necessary and sufficient to perform these functions is selected by the designer, and other SSCs that are capable of performing the same safety functions may be considered for their risk significance or if they are needed for DID.

4 High-consequence BDBEs are BDBEs whose consequences exceed the 10 CFR 50.34 dose limits.

expected to use the terminology in NEI 18-04 and as needed, identify exceptions to and exemptions needed from NRC regulations. If TerraPower does not comply with 10 CFR 50.2 as stated, they must request an exemption; however, neither compliance with, nor an exemption from, the definition of safety-related SSCs in 10 CFR 50.2 was discussed in the TR. This issue is discussed further in the Limitations and Conditions section of this SE.

3.0 TERRAPOWER REGULATORY ANALYSES 3.1 Analyses of 10 CFR 50.10 Pursuant to 10 CFR 50.10(c), [n]o person may begin the construction of a production or utilization facility on a site on which the facility is to be operated until that person has been issued either a CP or a limited work authorization [LWA]. The 10 CFR 50.10(a) definition of construction is divided in two parts: 10 CFR 50.10(a)(1) specifies activities deemed to constitute construction, and 10 CFR 50.10(a)(2) specifies activities which are excluded from the definition.

Paragraph 10 CFR 50.10(a)(1) specifies the following definition of construction:

(1) Activities constituting construction are the driving of piles, subsurface preparation, placement of backfill, concrete, or permanent retaining walls within an excavation, installation of foundations, or in-place assembly, erection, fabrication, or testing, which are for:

(i) Safety-related structures, systems, or components (SSCs) of a facility, as defined in 10 CFR 50.2; (ii) SSCs relied upon to mitigate accidents or transients or used in plant emergency operating procedures; (iii) SSCs whose failure could prevent safety-related SSCs from fulfilling their safety-related function; (iv) SSCs whose failure could cause a reactor scram or actuation of a safety-related system; (v) SSCs necessary to comply with 10 CFR part 73; (vi) SSCs necessary to comply with 10 CFR 50.48 and criterion 3 of 10 CFR part 50, appendix A; and (vii) Onsite emergency facilities, that is, technical support and operations support centers, necessary to comply with 10 CFR 50.47 and 10 CFR part 50, appendix E.

In Section 8.1 of the TR, TerraPower evaluated the EI SSCs against the criteria in 10 CFR 50.10(a)(1) to determine whether construction of EI SSCs would constitute construction under the regulation and would therefore require NRC authorization through a CP or LWA. The NRC staff notes that the analyses provided in the TR only relate to the 10 CFR 50.10 definition of construction and do not address definitions of construction elsewhere in NRC regulations (e.g., 10 CFR Part 51). This is highlighted as a limitation and condition in the Limitations and Conditions section of this SE.

10 CFR 50.10(a)(1)(i) - (iii)

TerraPower stated in the TR that, since the Natrium design includes only NST SSCs on the EI, does not rely on any NST SSCs to provide mitigation for an accident or transient, and does not have any NST SSCs whose failure could prevent SR SSCs from fulfilling their safety-related function, the 10 CFR 50.10(a)(1)(i) through (iii) criteria do not apply to SSCs on the EI.

Regarding 10 CFR 50.10(a)(1)(i), since the criterion applies only to SR SSCs the NRC staff finds it reasonable in principle to exclude NST SSCs (with no special treatment) from consideration. However, the regulation specifically states that construction includes safety-related SSCs of a facility as defined in 10 CFR 50.2. As noted previously in Section 2.0 of this SE, the TR does not address compliance with or exemption from the 10 CFR 50.2 definition of safety-related SSCs. If TerraPower seeks an exemption from the 10 CFR 50.2 definition of safety-related SSCs, TerraPower must also seek an exemption from 10 CFR 50.10(a)(1)(i). If compliance with the 10 CFR 50.2 definition is demonstrated, an exemption may not be needed.

This is highlighted in the Limitations and Conditions section of this SE.

With respect to 10 CFR 50.10(a)(1)(ii), TerraPower stated that the Natrium plant design does not rely on any NST SSC - which includes all SSCs on the EI, as discussed above - to provide mitigation for an accident or transient. The NRC staff considers this reasonable to partially address criterion (ii), provided that the EI SSCs remain appropriately categorized as NST as the design matures; this is addressed in the discussion provided in Section 2.0 of this SE and in the Limitations and Conditions section of this SE. TerraPowers logic is also consistent with the design and transient response of the reactor as discussed in SE Sections 1.1 through 1.8.

However, TerraPower did not address plant emergency operating procedures (EOPs), which are included in the regulation. Since the NRC staff has not yet reviewed EOPs for the Natrium facility, future licensing applications or correspondence referencing this TR must demonstrate that EI SSCs are not used in the EOPs to support a conclusion that no EI SSCs meet this aspect of the 10 CFR 50.10(a)(1)(ii) criterion; this is included in the Limitations and Conditions section of this SE.

TerraPower dispositioned 10 CFR 50.10(a)(1)(iii) by stating that there are no NST SSCs in the Natrium design whose failure could prevent SR SSCs from fulfilling their safety-related function.

This is reasonable, and the NRC staff finds it to be logically consistent with the definition of safety-related from NEI 18-04, in that any SSCs whose failure could prevent an SR SSC from fulfilling its safety-related function would not be characterized as NST under the risk-informed process. However, as stated in SE Section 2.0, the definition of safety-related SSCs in NEI 18-04 differs from the definition in 10 CFR 50.2. Compliance with, or an exemption from, the 10 CFR 50.2 definition is addressed in the Limitations and Conditions section of this SE.

10 CFR 50.10(a)(1)(v) - (vii)

TerraPower determined that criterion (v) is not applicable to EI SSCs. With respect to physical security program SSCs, TerraPower stated in the TR that: [c]onstruction activities for SSCs necessary to comply with 10 CFR 73 include the preparation and building of physical barriers and structures and associated hardware and detection systems for the physical security program. None of these physical security program SSCs are located on the EI for the Natrium reactor. With respect to cyber security program SSCs, TerraPower stated in the TR that no digital components or control systems identified as critical digital assets (CDAs) would be installed or activated prior to receipt of a CP. While the NRC staff finds this to be plausible based on the safety significance of SSCs on the EI, TerraPower did not provide detailed information on how the physical and cyber security programs would be implemented in such a way that they would deliberately avoid involving the EI. Thus, additional justification must be provided in future licensing applications or correspondence referencing this TR to support a conclusion that no EI SSCs meet the 10 CFR 50.10(a)(1)(v) criterion; this is included in the Limitations and Conditions section of this SE. The NRC staff also determined that, contrary to TerraPowers overall conclusion, criterion (v) is applicable to CDAs on the EI; however, TerraPowers proposed disposition of this issue, which is to defer installation of CDAs on the EI until after the CP is issued, is appropriate.

TerraPower determined that criterion (vi) is not applicable to EI SSCs because fires on the EI would not prevent the ability to achieve and maintain safe shutdown of the reactor. The NRC staff reviewed TerraPowers assessment of 10 CFR 50.10(a)(1)(iv) and determined that it was adequate based on the classification of the EI systems as NST and the NRC staffs review and conclusions documented in Section 1.0 of this SE that the Natrium reactor has the capability for safe shutdown using only NI SSCs.

TerraPower also evaluated criterion (vii) as not applicable to the EI because onsite emergency facilities are not located on the EI. This is consistent with plant facilities layouts provided to the NRC staff in public meetings (e.g., in [10]) and could reasonably be a design objective. However, as with criterion (v), additional justification must be provided in future licensing applications or correspondence referencing this TR to support a conclusion that no EI SSCs meet the 10 CFR 50.10(a)(1)(vii) criterion, as discussed in the Limitations and Conditions section of this SE.

10 CFR 50.10(a)(1)(iv)

Contrary to the other criteria, TerraPower determined that the 10 CFR 50.10(a)(1)(iv) criterion is applicable to SSCs on the EI. The NRC staff agrees with this assessment. The analysis provided by TerraPower in the TR indicates that failures in the TSS can cause a reactor scram, and this may be the case for other EI SSC failures if the failures are not resolved within a certain period (which is defined by the amount of thermal salt available in the cold salt tank). Since this criterion is applicable, TerraPower intends to submit a request for an exemption from 10 CFR 50.10(a)(1)(iv).

TerraPower stated in Section 8.2 of the TR that the rationale for requesting an exemption from 10 CFR 50.10(a)(1)(iv) would be the same as that for requesting an exemption from 10 CFR 50.65(b)(2)(iii). TerraPower based this determination on discussion in the LWA rule issuance (72 FR 57415), which stated that the criteria in 10 CFR 50.10(a)(1) were intended to encompass those SSCs which have a reasonable nexus to radiological health and safety or common defense and security. As further stated in the rule issuance, the NRC chose to base the criteria in 10 CFR 50.10(a)(1)(i) through (iv) on the scoping criteria used in 10 CFR 50.65(b) in part because [the]

definition is well understood and there is good agreement on its implementation. RG 1.206 [6],

additionally indicates that it is acceptable to apply guidance developed for the maintenance rule to determine which SSCs are within the scope of the definition of construction, for criteria (i) through (iv). While the NRC staff understands that an exemption from these regulations could have the same basis, due to the similar language and underlying basis of the regulations, no such exemption has yet been submitted. The NRC staff is not taking a position in this SE on any prospective exemption request the NRC might receive.

3.2 Analyses of 10 CFR 50.65 Section 50.65 of 10 CFR requires licensees to have a program that monitors the performance or condition of certain SSCs or demonstrates the performance or condition of these SSCs through appropriate preventative maintenance to provide reasonable assurance that they are capable of fulfilling their intended functions. The scope of SSCs that must be subject to this program is defined in 10 CFR 50.65(b):

(b) The scope of the monitoring program specified in paragraph (a)(1) of this section shall include safety related and nonsafety related [SSCs], as follows:

(1) Safety-related [SSCs] that are relied upon to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the guidelines in § 50.34(a)(1), § 50.67(b)(2), or § 100.11 of this chapter, as applicable.

(2) Non-safety related structures, systems, or components:

(i) That are relied upon to mitigate accidents or transients or are used in plant [EOPs]; or (ii) Whose failure could prevent safety-related [SSCs] from fulfilling their safety-related function; or (iii) Whose failure could cause a reactor scram or actuation of a safety-related system.

The requirements of 10 CFR 50.65(b)(1) are outside the scope of the NRC staffs evaluation of TerraPowers regulatory analysis, which focused on 10 CFR 50.65(b)(2).

TerraPower determined that criteria (i) and (ii) do not apply to SSCs located on the EI for the same reasons 10 CFR 50.10(a)(1)(ii) and (iii) were determined to be not applicable. The conclusions drawn by the NRC staff in SE Section 3.1 are applicable to these criteria as well.

TerraPower also stated in the TR that the Natrium reactor does not use any NST SSC in EOPs.

However, as previously discussed, the NRC staff has not yet reviewed EOPs for the Natrium facility; therefore, future licensing applications or correspondence referencing this TR must provide additional information to justify a conclusion that EI SSCs are not used in the EOPs, as listed in the Limitations and Conditions section of this SE.

TerraPower also determined that criterion (iii), which is similar to 10 CFR 50.10(a)(1)(iv), is applicable at least to the portion of the TSS located on the EI. Under this regulation, the portion of the TSS on the EI would be subject to the maintenance rule because a failure in this system would lead to a runback that, if failed, would cause a scram. The NRC staff agrees with this assessment, which is the same as that discussed in SE Section 3.1. Also, as discussed in Section 3.1 of this SE, TerraPower intends to seek an exemption from 10 CFR 50.65(b)(2)(iii). The NRC staff is not taking a position in this SE on any prospective exemption request the NRC might receive.

3.3 Analyses of 10 CFR Part 50, Appendix B Section 8.4 of the TR provides TerraPowers evaluation of 10 CFR Part 50, Appendix B, (hereafter referred to as Appendix B). TerraPower states that Appendix B provides quality assurance requirements for the design, manufacture, construction, and operation of SSCs, and that the requirements of Appendix B apply to all activities affecting the SR functions of those SSCs.

TerraPower determined that, since SSCs located on the EI will be classified as NST according to the NEI 18-04 process, they do not affect the SR functions of SSCs used for mitigation and are therefore not subject to the requirements of Appendix B.

The introduction to Appendix B states, in part:

Nuclear power plants include SSCs that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. This appendix establishes quality assurance requirements for the design, manufacture, construction, and operation of those SSCs. The pertinent requirements of this appendix apply to all activities affecting the safety-related functions of those SSCs; these activities include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying.

Accordingly, Appendix B is applicable to all activities affecting the safety-related functions of SSCs that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. The NRC staff considered that systems classified as NST under the NEI 18-04 process would not be involved in preventing or mitigating accidents, based on the risk-informed safety classification process, and would furthermore not be involved in supporting the SR functions of SSCs that are used in the prevention or mitigation of accidents.

However, as stated in SE Section 2.0, the definition of safety-related SSCs in NEI 18-04 differs from the definition in 10 CFR 50.2. Compliance with, or an exemption from, the 10 CFR 50.2 definition is addressed in the Limitations and Conditions section of this SE. Thus, the NRC staff determined that TerraPowers evaluation of the requirements of Appendix B, with respect to NST SSCs on the EI is acceptable, subject to this limitation and condition. The NRC staffs determination in this regard is based on TerraPowers summary of its preliminary implementation of the NEI 18-04 SSC classification process and is subject to confirmation; this is also documented in the Limitations and Conditions section of this SE.

3.4 Analyses of 10 CFR Part 55 TerraPower states in TR Section 8.2 that the Natrium design removes direct interaction between the nuclear reactor and the main turbine generator and that, due to the lack of direct interaction, operation of the main turbine generator is not an apparatus or mechanism whose manipulation directly affects the reactivity or power level of the reactor. TerraPower goes on to state that, on this basis, the Natrium design allows for a non-licensed individual to fully operate and control the main turbine generator.

The NRC staff evaluated TerraPowers regulatory analysis in TR Section 8.2 within the context of the regulations of 10 CFR Parts 50 and 55, the associated regulatory history, and relevant statutory requirements. The NRC staff notes that the AEA defines operators under Section 11 as individuals who manipulate the controls of utilization facilities and that it does not define what those controls consist of, thereby leaving that definition to be made by the NRC. The AEA further mandates under Section 107 that individuals who operate utilization facility controls must be licensed by the NRC.

From the inception of operator licensing in 1956 (21 FR 359), manipulation of the controls of a utilization facility was restricted to licensed operators under 10 CFR 50.54(i). This specific regulation is closely linked to the AEA Section 107 mandate that is discussed above. The original 1956 definition of controls (21 FR 6) was very broad and encompassed mechanisms which by manipulation or failure to manipulate singly or in combination could result in the release of atomic energy or radioactive material in amounts determined by the Commission to be sufficient to cause danger to the health and safety of the public.

However, in 1963 (28 FR 3197), the agency narrowed this definition of controls, stating that

[t]his narrower interpretation is in accord with the original Commission intent. This amended definition of controls (which remains unchanged to the present day) is limited in scope to apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor. A separate 1963 rulemaking (28 FR 3196) also introduced 10 CFR 50.54(j) which, in contrast with 10 CFR 50.54(i), does not limit the manipulation of apparatus and mechanisms other than controls to performance by licensed operators and senior operators, but states, Apparatus and mechanisms other than controls, the operation of which may affect the reactivity or power level of a reactor shall be manipulated only with the knowledge and consent of an operator or senior operator licensed pursuant to part 55 of this chapter present at the controls.

Thus, the regulations recognize a distinction between an apparatus or mechanism whose manipulation directly affects the reactivity or power level of the reactor, and therefore is a control, and a non-control apparatus or mechanism whose operation may affect the reactivity or power level of the reactor and is subject to 10 CFR 50.54(j).

Thus, the word direct, as used in the definition of controls, is central to understanding the meaning of the related requirements of 10 CFR 50.54(i) and (j). In analyzing this, the NRC staff recognized that substituting wording of equivalent meaning for direct yielded a working interpretation of these requirements that was suitable for use in evaluating the TR. Specifically, "controls" can be interpreted to mean apparatus and mechanisms that, when manipulated, affect the reactor power level or reactivity without also needing something intermediate to make that happen. Manipulations of this type fall under the scope of 10 CFR 50.54(i) and their performance is restricted to licensed operators and senior operators. Thus, the presence, or absence, of a significant intermediary between any given manipulation and the reactivity or power level effects on the reactor is the key factor that the NRC staff, in its judgement, identified as being the essential determinant of whether that manipulation falls under the scope of 10 CFR 50.54(i).

The NRC staff note this overall characterization is consistent with the Commission perspective expressed under 28 FR 3197 which disfavored an excessively broad definition of what constitutes the controls of a facility.

The NRC staff noted that the TSS provides thermal energy storage capacity equivalent to several hours of full electrical generation, such that reactor power is not directly correlated to EI steam generation over significant timeframes. The Natrium design thus enables the reactor and EI steam loads (such as a turbine used for electric power generation) to operate at power levels that are different from one another. The NRC staff evaluated the implications of this design configuration and determined that the TSS acts as a significant intermediary between manipulations involving EI steam loads and reactivity effects on the reactor. Based upon this, the NRC staff concluded that manipulations of Natrium apparatus and mechanisms that affect EI steam loads do not directly affect the reactivity or power level of the reactor and, therefore, do not fall under the scope of 10 CFR 50.54(i). TerraPower did not discuss compliance with 10 CFR 50.54(j) in its TR; the NRC staff addresses this regulation in the Limitations and Conditions section of this SE.

LIMITATIONS AND CONDITIONS The NRC staff identified the following limitations and conditions, applicable to any licensee or applicant referencing this TR:

1. The NRC staffs review identified key aspects of the Natrium plant design and transient response that play a significant role in the independence of the Natrium NI and EI, as discussed in Sections 1.1 through 1.8 of this SE. Applicants or licensees referencing this TR must propose a Natrium design with these key aspects or justify that any departures from these key aspects do not affect the conclusions in the TR and this SE.

2. The conclusions reached in this SE are not valid if a process other than that described in NEI 18-04 is used to perform the safety classification or to the extent that any SSCs on the Natrium EI are found to have a safety classification other than NST. Thus, applicants or licensees referencing this TR must confirm, with appropriate justification, that all EI SSCs are classified as NST using NEI 18-04. Such justification must provide, or reference and make available for audit, the analyses supporting EI SSC classifications.

3. The TR does not address differences between the 10 CFR 50.2 and NEI 18-04 definitions of safety-related SSCs and does not discuss whether TerraPower plans to demonstrate compliance with the 10 CFR 50.2 definition or seek an exemption.

Applicants or licensees referencing this TR must comply with the 10 CFR 50.2 definition or propose an exemption. If an applicant or licensee referencing this TR proposes an exemption from 10 CFR 50.2, an exemption to 10 CFR 50.10(a)(1)(i) must also be proposed, since that regulation references the 10 CFR 50.2 definition.

4. The NRC staffs review of what constitutes construction is limited to the provisions of 10 CFR 50.10 and does not address other definitions of construction (e.g., those in the environmental regulations in 10 CFR Part 51 or other non-NRC regulations).

5. Applicants or licensees using this TR as a basis for non-applicability of 10 CFR 50.10(a)(1)(ii) or 10 CFR 50.65(b)(2)(i) to the EI must provide additional information demonstrating that EI SSCs are not used in the EOPs.

6. Applicants or licensees using this TR as a basis for non-applicability of 10 CFR 50.10(a)(1)(v) to the EI must provide detailed information demonstrating that the physical and cyber security programs would be implemented in such a way that the EI would not include any SSCs that meet the 10 CFR 50.10(a)(1)(v) criterion (i.e., not include any SSCs necessary to comply with 10 CFR part 73, or otherwise not construct the EI SSCs necessary to comply with 10 CFR Part 73 until a CP has been issued).

7. TerraPower asserted in the TR that onsite emergency facilities necessary to comply with 10 CFR 50.47 or 10 CFR Part 50 Appendix E and facilities for providing onsite emergency first aid and decontamination are not located on the EI. Applicants or licensees using this TR as a basis for non-applicability of 10 CFR 50.10(a)(1)(vii) must provide detailed information to demonstrate this assertion.

8. This TR does not address the requirements of 10 CFR 50.54(j), and the NRC staff does not provide any evaluation in this SE of the implications of the Natrium design as it relates to this specific regulation. Thus, any Natrium facility licensee or applicant for an operating license (OL) or combined license (COL) referencing this TR must, in the absence of receiving an exemption, ensure that manipulation of any EI apparatus or mechanism which may affect the reactivity or power level of the reactor is only permitted to occur with the knowledge and consent of a licensed operator or senior operator.

9. Under 10 CFR 55.31(a)(5), reactivity manipulations for operator license applicant experience requirements must involve operating controls which, as discussed in Section 3.4 of this SE, are associated with direct reactivity or power changes.

Therefore, any apparatus or mechanism determined to not be a control must, logically, also be excluded from being acceptable for applicant experience credit under 10 CFR 55.31(a)(5). Consistent with this, applicants for operator and senior operator licenses at a Natrium facility where the facility licensee (or applicant for an OL or COL) references this TR cannot rely upon the manipulation of apparatus and mechanisms that affect EI steam loads (to include a turbine used for electrical generation) for the purposes of satisfying the operator license applicant experience requirements of 10 CFR 55.31(a)(5).

10. The NRC staff is not taking a position in this SE on any prospective exemption request that the NRC might receive on matters discussed in the TR.

CONCLUSION This TR provides an evaluation of the NRC regulations pertaining to the design interface between the Natrium NI and EI and includes a supporting summary of key aspects of the Natrium reactor plant design, interfaces, safety features, and basic plant transient analysis, and TerraPowers process for classifying SSCs. The NRC staff reviewed the key aspects and TerraPowers evaluation of select regulatory requirements and concluded that the TR is acceptable for use in licensing applications subject to the Limitations and Conditions in this SE.

REFERENCES

[1] R. Sprengel, letter to U.S. Nuclear Regulatory Commission, Submittal of TerraPower, LLC Topical Report, Regulatory Management of Natrium Nuclear Island and Energy Island Design Interfaces, dated Oct. 04, 2022 (ML22277A824).

[2] M. Sutton, email to G. Wilson, TerraPower: Completeness Determination for the Natrium Nuclear Island and Energy Island Design Interfaces Topical Report, dated Nov. 17, 2022 (ML22319A153).

[3] U.S. Nuclear Regulatory Commission, NRC 898 - Topical Report Completeness Determination for TerraPower, LLC Natrium Nuclear Island and Energy Island Topical Report, dated Nov. 16, 2022 (ML22319A131).

[4] M. Sutton, Regulatory Management of Natrium Nuclear Island and Energy Island Design Interfaces Audit Plan, dated Jan. 05, 2023 (ML22353A642).

[5] M. Sutton, email to G. Wilson, "TerraPower: Natrium EI-NI Separation TR Audit Summary,"

dated June 16, 2023 (ML23167A478).

[6] U.S. Nuclear Regulatory Commission, Applications for Nuclear Power Plants, Revision 1, RG 1.206, dated Oct. 3, 2018 (ML18131A181).

[7] U.S. Nuclear Regulatory Commission, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4, Regulatory Guide 1.160, dated Aug. 31, 2018 (ML18220B281).

[8] U.S. Nuclear Regulatory Commission, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, Regulatory Guide 1.233, Revision 0, dated June 30, 2020 (ML20091L698).

[9] Nuclear Energy Institute, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1, NEI 18-04, dated Aug. 29, 2019 (ML19241A472).

[10] R. Sprengel, letter to the U.S. Nuclear Regulatory Commission, TerraPower Plant and Licensing Strategy Overview Presentations to the Advisory Comittee on Reactor Safeguards, dated Mar. 31, 2023 (ML23090A228).

Project Manager: Mallecia Sutton, NRR Principal Contributors: Reed Anzalone, NRR Jesse Seymour, NRR Ben Adams, NRR