SECY-24-0008, Enclosure 1 - Technical, Licensing, and Policy Considerations for Factory-Fabricated Micro-Reactors

SECY-24-0008: Enclosure 1 - Technical, Licensing, and Policy Considerations for Factory-Fabricated Micro-Reactors

ML23207A251
Person / Time
Issue date:	01/24/2024
From:	Dan Dorman NRC/EDO
To:	Commissioners NRC
Kennedy W
Shared Package
ML23207A252	List: SECY-24-0008, Enclosure 1 - Technical, Licensing, and Policy Considerations for Factory-Fabricated Micro-Reactors
References
SECY-24-0008
Download: ML23207A251 (31)

Text

Technical, Licensing, and Policy Considerations for Factory-Fabricated Micro-Reactors

This enclosure includes various topics related to the licensing and deployment of factory-fabricated micro-reactors (and potentially applicable to other new reactors, as appropriate).

Some of these are topics raised by developers through formal pr e-application engagement with the U.S. Nuclear Regulatory Commission (NRC) staff and in other interactions, such as the periodic Advanced Reactor Stak eholder Meetings organized by the NRC staff. Some of the topics have previously been considered in SECY-20-0093, Policy and Licensing Considerations Related to Micro-Reactors, dated October 6, 2020 (Agencywide Documents Access and Management System Accession No. ML20129J985), and in the context of small modular reactors or non-light-water reactors but are revisited here wit h the attributes of factory-fabricated micro-reactors and related deployment models in mind. The NRC s taff will address design-specific issues on a case-by-case basis.

This enclosure includes the NRC staffs near-term strategies an d next steps for addressing each topic. The near-term strategies provide means for the NRC staff to address each topic under the existing regulatory framework without additional Commission dir ection. These are included to inform the Commission of approaches that are available now. The next steps focus on longer-term approaches, and the NRC staff will engage the Commission o n any future policy topics for factory-fabricated micro-reactors, including any related to saf ety, security, emergency preparedness, and environmental reviews.

1. Timeframe for Authorization to Operate at the Deployment Sit e

Deployment Model Considerations

Factory-fabricated micro-reactors may have significantly simpler and shorter duration construction activities at the deployment site than large light -water reactors, which typically take several years to construct. A key aspect of factory-fabricated micro-reactor deployment models is the ability to move a factory-fabricated module from the fac tory to the deployment site and place it into operation as a nuclear power plant in a much shor ter time than it takes to construct a large light-water reactor at the intended site of operation. Factory-fabricated micro-reactors that are of a self-contained design with the nuclear and bala nce-of-plant systems in one or a few containers that are fully fabricated at the factory would l ikely require only simple construction activities at the deployment site (e.g., pouring a small concrete pad on which to place the container housing the reactor). Factory-fabricated mi cro-reactor designs that have a core module design would likely require more complex constructi on activities at the deployment site (but still much simpler than construction activities for l arge light-water reactors), such as erecting a reactor building and installing power conversion equ ipment. Developers have suggested that self-contained micro-reactors might be ready for operation within days to weeks of beginning construction at the deployment site and micro-reac tors with a core module design might be ready within a few months.

Factory-fabricated micro-reactors may be licensed at the deploy ment site under either Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, or 10 CFR Part 52, Licenses, Certific ations, and Approvals for Nuclear Power Plants. In both cases, a mandatory hearing would be held, and a contested hearing opportunity would be provided, as part of the process for issui ng the authorization to construct the reactor at the deployment site. Factory-fabricated micro-re actors would likely have standardized designs that may be described in a referenced 10 C FR Part 52 design certification 2

or manufacturing license, which c ould reduce the scope of NRC review for deployment site licensing. Also, any final safety findings on final design info rmation in a construction permit application would be incorporated in the permit in accordance w ith 10 CFR 50.35(b) and subject to the backfitting requirements in 10 CFR 50.109, Backfitting.

Under a 10 CFR Part 50 approach, AEA section 189a.(1)(A) requir es an opportunity for a contested hearing (but not a mandatory hearing) in conjunction with issuance of the facility operating license. The regulations in 10 CFR 2.309(b)(3) provid e a minimum 60-day opportunity to request a hearing. The potential scope of such a hearing wou ld be the entirety of the operating license application, but the hearing scope would be r educed to the extent that the operating license application references an earlier NRC license or approval providing finality on the matters resolved therein, such as a manufacturing license. Subparts C and L and Appendix B to 10 CFR Part 2, Agency Rules of Practice and Procedure, p rovide the rules of general applicability, procedures, and model milestones for such a hear ing. If no hearing is requested, the Commission could issue the operating license immediately up on the closure of the 60-day hearing request period, provided all other requirements are met.1 If a hearing is requested, the issuance of the license would have to wait until a presiding of ficer decision resolving the contested issues, which might occur after a hearing (if the req uest is granted) or a presiding officer decision on the hearing request (if the request is deni ed), either of which could take many months according to the procedur es and model milestones in 10 C FR Part 2.

The environmental review may also affect the timeframe for depl oyment under the 10 CFR Part 50 licensing process. After issuing a permit to construct a nuclear power reactor with a supporting environmental impact statement (EIS), the regulation s in 10 CFR 51.20(b)(2) and 10 CFR 51.95(b) require that the NRC staff publish a supplement to the construction permit EIS to support issuance of the operating license. The process of pr eparing and issuing the supplement to the EIS includes publication of a draft supplemen t with an additional period for public comment, and various consultations with external stakeho lders. In addition, the National Environmental Policy Act of 1969, as amended (NEPA), was amende d in June 2023 to include a new requirement to complete EISs within 2 years. 2 Recent improvements in the environmental review process along with standardized, relatively simple react or designs, and small reactor sites could reduce the time to complete a supplement to an EIS to less than 24 months, with further process improvements planned. This is in contrast to th e environmental review for a combined license, in which the environmental review is complete d before license issuance and no additional environmental review is required in connection wi th the 10 CFR 52.103(g) finding.

Under a 10 CFR Part 52 combined license, an opportunity for hea ring is required by 10 CFR 52.103(a) and AEA section 189a.(1)(B)(i) on whether the facility as constructed complies, or on completion will comply, with the acceptance cri teria of the [combined] license.

Both 10 CFR 52.103(a) and AEA section 189a.(1)(B)(i) require th e NRC to publish a notice of intended operation providing this hearing opportunity at least 180 days before the scheduled date for initial loading of fuel by the combined license holder, with a 60-day period for the opportunity to request a hearing. The potential scope of such a hearing would be limited to the inspections, tests, analyses, and acceptance criteria (ITAAC) i ncluded in the combined license.

1 AEA section 189a.(1)(A) states, in part, that [i]n cases where such a construction permit has been issued following the holding of such a hearing, the Commission may, in the absence of a request therefor by any person whose interest may be affected, issue an operating license wi thout a hearing, but upon thirty days notice and publication once in the Federal Register of its intent to do so (emphasis added).

2 Fiscal Responsibility Act of 2023, Public Law No. 118-5, § 321(b). The 2-year deadline is in the new NEPA Section 107(g).

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NRC regulations in 10 CFR Part 52, Subpart C include additional timing requirements written in terms of the initial loading of fuel under a combined license:

• Under 10 CFR 52.103, Operation under a combined license, the licensee shall notify the NRC of its scheduled date for initial loading of fuel no la ter than 270 days before the scheduled date.

• The regulations in 10 CFR 52.99, Inspection during constructi on, include requirements on the timing of licensee notifications of ITAAC closure and co mpletion and NRC publication of related notices. In particular, 10 CFR 52.99(c)( 3) states that if the licensee has not provided an ITAAC closure notification under 10 CFR 52. 99(c)(1) for all ITAAC by 225 days before the scheduled date for initial loading of fu el, then the licensee must provide an uncompleted ITAAC notification no later than 225 day s before the scheduled initial fuel load date to describe how the licensee will comple te the uncompleted ITAAC.

The purpose of this requirement, in part, is to ensure that inf ormation related to ITAAC closure is available to the NRC staff and the public at the tim e the Commission publishes the 60-day notice of opportunity for hearing in the Federal Register as required by AEA section 189a.(1)(B)(i).

These regulations and the final ITAAC hearing procedures publis hed on July 1, 2016 (Volume 81 of the Federal Register (FR), page 43266 (81 FR 43266)) were developed to ensure the Commission will meet the requirements of AEA section 189a.(1)(B )(v) and 10 CFR 52.103(e),

which provide that the Commission shall, to the maximum possibl e extent, decide on issues raised by the hearing request within 180 days of the publicatio n of the notice of intended operation or the anticipated date for initial loading of fuel i nto the reactor, whichever is later.

For large light-water reactors, the timeframes for licensee not ifications and Commission actions required by AEA section 189 and Subpart C of 10 CFR Part 52 fit within the overall construction schedule, which is usually several years. For factory-fabricate d micro-reactors with much shorter deployment schedule goals, these timeframes could resul t in delays in entering the reactor into operation. If the licensee were to notify the Comm ission of the intended date of initial fuel load upon receipt of the combined license, that no tification would start the 270-day period. By the end of that 270-day period, the Commission would normally complete the hearing, if requested and granted, and be able to determine whe ther it can make the 10 CFR 52.103(g) finding. Based on the 270-day notification by the licensee, the Commission would have 90 days to publish the notice of intended operation, which triggers the 60-day opportunity to request a hearing required by AEA section 189a.( 1)(B)(i). However, as discussed in the NRCs ITAAC hearing procedures, the Commission has estab lished a goal to publish the notice of intended operation at least 210 days before scheduled fuel load.

The Commission has previously established that a licensee may u nder certain conditions begin operation before the scheduled date of initial fuel loading sub mitted to the Commission under 10 CFR 52.103(a). In the Federal Register publication of the final ITAAC hearing procedures, the NRC stated (81 FR at 43273) that the licensee can, consistent with 10 CFR 52.103(a),

move up its scheduled fuel load date after the notice of intend ed operation is published. Such a contraction in the licensees fuel load schedule would have no effect on the hearing schedule, but as a practical matter, the NRC would consider such a contra ction in the licensees schedule as part of its process for making the 10 CFR 52.103(g) finding and the adequate protection determination for interim operation. In response to comments o n the proposed ITAAC hearing procedures, the NRC stated that in the absence of a hearing or if the hearing issues are resolved early in favor of the licensee, the licensee will be a llowed to operate if and after the 4

10 CFR 52.103(g) finding is made. 3 The NRC also stated that if a hearing is held and has not been completed, but the NRC staff has made the 10 CFR 52.103(g) finding and the Commission has made the adequate protection determination for i nterim operation, then the licensee will be allowed to enter into interim operation.

Near Term Strategy

The NRC staff intends to use the existing regulations in 10 CFR Parts 2, 50, and 52 in connection with issuing operating licenses and authorizing oper ation under a combined license.

The NRC staff also intends to use the existing final ITAAC hear ing procedures.

Several steps may be taken to potentially shorten the timeframe for deployment of a factory-fabricated micro-reactor under a combined license. A key strate gy would be to publish the notice of intended operation as early as possible. The NRC coul d not publish this notice before combined license issuance because AEA section 189a.(1)(B)(i)-(i i) provides that the hearing opportunity is on conformance with the acceptance criteria in the combined license. However, a licensee could provide the 10 CFR 52.103(a) notification of its scheduled date for initial fuel load4 and either an ITAAC closure notification or an uncompleted ITA AC notification for all ITAAC immediately upon receipt of the combined license. If the combined license applicant intends to do this, it should inform the NRC staff of its inten tion in the combined license application or by other means so that the NRC staff can make ne cessary arrangements to prepare the notice of intended operation. Consistent with the N RCs experience with the ITAAC proceedings for Vogtle Electric Generating Plant, Units 3 and 4, the NRC could make the ITAAC notifications publicly available and publish the notice of inte nded operation within about 15 days after receipt of the ITAAC notifications.

During the 60-day opportunity to request a hearing, the license e would presumably complete construction of the reactor, provide the appropriate notificati ons related to ITAAC required by 10 CFR 52.99(c), and notify the NRC of its update to the date s cheduled for initial fuel load in accordance with 10 CFR 52.103(a). If the NRC staff is able to c onclude that all acceptance criteria are met by the close of the 60-day period for requesti ng a hearing and no hearing was requested, the NRC staff would aim to make the 10 CFR 52.103(g) finding shortly thereafter, possibly within 5 days. This could result in a deployment timef rame of as little as about 80 days after combined license issuance, w hich may be driven primarily by the statutory requirement in AEA section 189a.(1)(B)(i) that the notice of intended operatio n provide that any person whose interest may be affected by operation of the plant, may within 60 days request the Commission to hold a hearing. When a hearing is requested, the minimum t imeframe would be extended in accordance with the final ITAAC hearing procedures. That is, the 10 CFR 52.103(g) finding might be issued (1) after the Commissions decision on the hear ing request if the request is denied, (2) after a decision allowing interim operation if the hearing request is granted and the requirements for interim operation are met, or otherwise, (3) a fter the presiding officer has issued the decision after the hearing.

3 Comment Summary Report - Procedures for Conducting Hearings on Whether Acceptance Criteria in Combined Licenses Are Met, Section 5.G (ML16167A464).

4 If fuel is loaded at the manufacturing facility, then the licensee would not be able to notify the NRC of its scheduled date for initial loading of fuel as required by 10 CFR 52.103(a) because the manufacturer would be loading fuel at its site under its license rather than the licensee for the deployment site loading fuel at the deployment site under its license. In that circumstance, the licensee could alternatively provide a schedule for the removal of the features to preclude criticality proposed by the NRC staff in this paper.

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If a hearing is requested, the ITAAC hearing procedures allow t he licensee and the NRC staff 25 days to answer the hearing request and establish a milestone of 30 days after the answers for a Commission ruling on the hearing request. If the Commissi on does not grant a hearing request, this would add 55 days to the minimum deployment timef rame compared to the scenario in which no hearing request is filed (135 days total). If the Commission does grant the hearing request, then the minimum timeframe could be extended b y an additional 70 to 94 days (205 to 229 days total), although interim operation may be allo wed during the hearing if the Commission makes the adequate protection determination for inte rim operation and the NRC staff is able to make the 10 CFR 52.103(g) finding.

Under 10 CFR Part 50, the Commission would notice the opportuni ty for hearing in conjunction with its notice docketing the application for the operating lic ense for the deployment site. The NRC staff could then complete its final safety evaluation durin g the 60-day period of the opportunity to request a hearing, assuming that the final desig n, site-specific issues, technical specifications, and operational programs have been reviewed and approved during the construction permit proceedings or other prior approvals (e.g., topical reports) and the operating license application does not introduce any departures. If a hea ring is not requested and all other requirements are met, the deployment timeframe could be shorten ed to approximately 95 days, which allows for 30 days to perform an acceptance review and do cket the application, 5 days to publish the notice of opportunity for hearing, and 60 days for the opportunity to request a hearing. If a hearing is requested, the timeframe will be exten ded in accordance with the regulations in 10 CFR Part 2. However, under the 10 CFR Part 50 process for issuing a facility operating license, the record of decision cannot be issued unti l both the safety and environmental reviews are completed. Characteristics of factory -fabricated micro-reactors, such as standardized designs and relatively small site footprints wi th limited construction activities at the deployment site, may allow for the NRC staff to complete th e required supplement to the EIS in less than the 24 months normally allotted under the curr ent process.5

Despite the differences in the requirements for issuance of an operating license or making the 10 CFR 52.103(g) finding, the NRC staff has identified the time needed to issue an EIS supplement to support issuance of an operating license as the m ain reason why the 10 CFR Part 50 process would result in a significantly different deplo yment schedule compared to 10 CFR Part 52. Ultimately, it will be up to an applicant to co nsider the differences in the two licensing processes and decide which one is better suited to it s particular deployment model.

Next Steps

The NRC staff intends to further assess the final ITAAC hearing procedures and the requirements in 10 CFR Part 2 based on (1) Commission direction on the options presented in this paper for features to preclude criticality, fuel loading, and operational testing; (2) the characteristics of factory-fabricated micro-reactors; and (3) f urther stakeholder input. If warranted, the NRC staff will propose an update to the final IT AAC hearing procedures for Commission consideration and consider rulemaking options, as ap propriate, to better tailor microreactor hearing processes to the characteristics of, and l icensing strategies for, micro-reactors.

In accordance with Staff Requirements Memorandum (SRM)-SECY 0001, Staff RequirementsSECY-21-0001Rulemaking PlanTransforming the NRC s Environmental

5 The Fiscal Responsibility Act of 2023 codifies a one-year deadline for completing environmental assessments and a 2-year deadline for completing EISs.

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Review Process, dated April 19, 2022 (ML22109A171), after comp leting several environmental reviews for advanced reactors, the NRC staff could further expl ore the idea of preparing environmental assessments to meet NEPA requirements for some ca tegories and subcategories of license applications currently falling within the scope of 10 CFR 51.20(b), and present options to the Commission. If such a proposal is develo ped and approved, environmental assessments could be used to determine whether an EIS was required, both for the construction permit and the operating license at the deploy ment site, which would require appropriate changes to 10 CFR Part 51, Environmental Protectio n Regulations for Domestic Licensing and Related Regulatory Functions, or exemptions. The use of environmental assessments to determine whether an EIS is necessary, could, pr ovided the environmental assessment results in a finding of no significant impact, substantially shorten the timeline such that the deployment timeframe could be dictated by the operatin g license contested hearing process.

2. Replacement of Factory-Fabricated Modules at the Deployment Site

Deployment Model Considerations

Factory-fabricated micro-reactor deployment models might includ e periodically removing factory-fabricated modules from the deployment site at the end of their operational lives or fuel cycles and replacing them with modules of the same design. This could involve shipping a factory-fabricated module away from the deployment site for ref ueling and refurbishment and then returning it to the deployment site or shipping a new modu le to the deployment site to replace the existing one. The replacement of factory-fabricated modules could result in the need for the deployment site licensee to have multiple fueled module s on site at some times to allow for transition from the operating module to the replacement mod ule with minimal downtime.

The NRC staff has considered potential licensing strategies for replacement modules under the current regulatory framework. The NRC staff previously addresse d licensing options for multi-module facilities in SECY-11-0079, License Structure for Multi -module Facilities Related to Small Modular Nuclear Power Reactors, dated June 12, 2011 (ML110620459). A key difference from the small modular reactors considered in SECY-1 1-0079 is that a replacement factory-fabricated micro-reactor module is not intended to be o perated at the same time as the module it is intended to replace. Each module would operate for a limited time (generally less than 10 years) and then would be removed from service and repla ced. Upon removal of the module from service, the deployment site licensee would install features to preclude criticality to take the module out of operation and store it on site before de commissioning or shipment to a decommissioning facility or a refurbishment and refueling facil ity.

The NRC staff has identified a licensing strategy under 10 CFR Part 52 that would have the initial combined license application include one final safety a nalysis report to address the requirements of 10 CFR 52.79, Contents of applications; techni cal information in final safety analysis report, for all factory-fabricated modules, including replacement modules, anticipated to be operated at the deployment site. 6 The application would also specify the number of

6 The NRC staff is aware that the designs of factory-fabricated m icro-reactors, as described in the FSAR for the manufacturing license, could evolve over the lifetime of a single module such that when the time comes to replace a module, the replacement may be of a different design. Under 10 CFR 52.171(b)(2), the deployment site licensee will need to amend its existing permits and licenses or seek new ones to account for the new module design described in the manufacturing license FSAR. Also, the deployment site licensee may include a module-specific appendix to the final safety analysis report that describes and analyzes the design differences between each module of a new design and the "standard" module.

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modules that could be present at the site simultaneously and op erated simultaneously. The combined license application would also include any permanently installed site-specific features such as structures or power conversion systems, if applicable. Under this approach, each module would receive its own combined license; however, the lic enses would be issued concurrently. This would be similar to Alternative 3: Individu al Reactor Module Licenses described in SECY-11-0079, which was the NRC staffs preferred alternative as the best approach for the licensing of multi-module power reactor facili ties. As noted in SECY-11-0079

Consistent with NRC regulations a nd existing practice, a [combined license]

application related to multiple modules at a single facility ca n undergo a single license review, safety evaluation report (SER), and hearing if a single license application is made for modules of essentially the same design. The precedent for this process comes from recent large light-water reactor [c ombined license]

applications that have been filed under 10 CFR Part 52 for two units (e.g., Vogtle Electric Generating Plant), and many [construction permits] and [operating licenses] issued under 10 CFR Part 50.

NRC regulations related to ITAAC (10 CFR 52.103(g)) adequately address the transition from construction to operation under 10 CFR Part 52 by allowing separate findings for each module. The individual license for e ach module would also support the transition from construction to operation unde r 10 CFR Part 50 by allowing the issuance of separate [operating licenses] at di fferent times for each module (which has been the historical practice for [constr uction permits]

issued for multiunit sites).

The EIS and hearing(s) required for combined license issuance w ould address all the factory-fabricated modules to be licensed to operate over the l ife of the deployment site. The licensee would be required to show that all ITAAC have been met and the Commission would need to issue its finding under 10 CFR 52.103(g) (and offer the associated opportunity for hearing) before the first module and every subsequent replaceme nt module would be authorized to begin operation under its combined license. There are potential timing impacts associated with completing ITAAC and potential ITAAC hearings f or each replacement module, but the license application, safety review, mandatory hearing, and opportunity for contested hearing on the combined license issuance would be conducted onl y once for all combined licenses for the modules anticipated to be operated at the faci lity. The NRC staff expects that licensees would know in advance of the need for the replacement module and therefore could plan accordingly to minimize potential impacts on plant downtim e caused by the time required for ITAAC completion, a potential ITAAC hearing, and the Commis sions finding required by 10 CFR 52.103(g).

For example, depending on the design of the factory-fabricated module and the installation process, the licensee could bring the replacement module on sit e, provide appropriate ITAAC notifications to the NRC, and complete the ITAAC hearing proces s well before the currently operating module reached the end of its life or fuel cycle. If timed appropriately, this could provide time for the Commission to make the finding required by 10 CFR 52.103(g) before the licensee removed the currently operating module from service. 7 Depending on the deployment facility and the licensees plans for operation, the final safe ty analysis report may need to

7 This scenario assumes that the licensee could complete all ITAAC for the replacement module while the existing module is still operating (e.g., the replacement module is installed at an onsite location different from the existing modules location).

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account for additional operating modules at the site, or condit ions in the combined license for the replacement module may need to specify that the replacement module would not be placed into operation unless site-specific conditions are met (e.g., a maximum number of modules operating simultaneously).

Under 10 CFR Part 50, the NRC could issue one construction perm it covering the construction of the facility (i.e., one or more factory-fabricated modules a nd the onsite balance-of-plant) including all replacement modules expected to be deployed at th e site. Issuance of the construction permit would involve a mandatory hearing, an oppor tunity for a contested hearing on the construction permit issuance, and an EIS that would cons ider all of the modules expected to be deployed at the site. The NRC could then issue s eparate operating licenses to authorize operation of each replacement module after each one i s installed at the deployment site. To the extent that the first module and all replacements would have the same standard design and any permanent onsite structures or features would ha ve been approved with the issuance of the operating license for the first module at the d eployment site, the NRC staff could leverage the safety review that had already been completed for subsequent operating licenses.

Issuance of the operating license would require the NRC to prov ide an opportunity for the public to request a hearing and to publish a supplement to the final E IS for the construction permit as required by 10 CFR 51.20(b)(2) and 10 CFR 51.95(b). As stated i n 10 CFR 51.95(b), the supplement to the EIS would cover only matters that differ from the final EIS for the construction permit or that reflect significant new information concerning m atters discussed in that final EIS.

Under the 10 CFR Part 50 approach, the NRC staff expects that, depending on the design of the factory-fabricated module, the potential downtime could be minimized by the operating license application being submitted well in advance of removing a currently operating module from service, which should allow the EIS to be supplemented and the contested hearing process to be completed. However, new issues might arise in the operating license process for a replacement module. This is in contrast to the approach using 10 CFR Part 52 under which the design and environmental reviews are completed and conclusi vely resolved upon issuance of the combined licenses.

Near Term Strategy

The NRC staff intends to use the existing regulations in 10 CFR Part 50 and 10 CFR Part 52, as informed by SECY-11-0079 and the approach described above, for licensing replacement of factory-fabricated modules. Factory-fabricated-micro-reactor de velopers and potential applicants will need to consider factors such as the number of modules that will be on site at one time, the expected operational states of the onsite modules, replacement frequency, and others when deciding whether to apply for licensing of replacem ent modules under 10 CFR Part 50 or 10 CFR Part 52.

Next Steps

The NRC staff intends to continue to engage in pre-application and stakeholder engagement activities to further understand and assess planned deployment models, the potential to streamline licensing pathways, and the need for additional guid ance under the current regulatory framework. The NRC staff will also consider whether other licensing approaches for multi-module sites described in SECY-11-0079, such as a single license for multiple modules or the master facility license alternative, could provide effici encies for licensing replacement modules at the deployment site. The NRC staff will continue to monitor developers and potential applicants plans to assess the viability of proposed strategies and any potential policy 9

issues needing further Commission engagement. If the NRC staff determines that alternative licensing strategies would require rulemaking or a policy decis ion, this would be addressed in a separate vote paper to seek Commission direction.

3. Autonomous Operation and Remote Operation

Deployment Model Considerations

During recent pre-application interactions with the NRC staff, micro-reactor developers expressed significant interest in the inclusion of autonomous 8 and remote operational characteristics in their proposed designs. A remote operational model tends to center around minimizing the numbers of both operators and other categories o f facility staff at the facility site, while an autonomous operational model seeks to eliminate relian ce upon the use of operators.

For the purposes of this paper, autonomous systems are consider ed those able to perform their task and achieve their functions independently (of the human op erator), perform well under significant uncertainties for extended periods of time with lim ited or nonexistent communication, with the ability to compensate for failures, all without extern al intervention.9 In the case of remote operations, the objective is to relocate staff to a cent ralized location and operate some reactors remotely. Such approaches differ dramatically from the current paradigm of commercial nuclear plant operations in which operators are required by 10 CFR 50.54(k) and (m) to maintain a continual onsite presence in control rooms. Addition ally, some operational approaches for very simple micro-reactor designs may not includ e a traditional control room, thereby requiring the evaluation of requested exemptions from r elevant regulations, such as 10 CFR 50.34(f)(2)(iii).10 Thus, developer-proposed deployment models that would involve relocating operators offsite or eliminating them entirely prese nt significant differences from traditional reactor designs and licensing paradigms.

As previously noted in SECY-20-0093, both autonomous and remote operations raise potential policy-related matters. For example, autonomous operation would entail reactivity manipulations being performed by automation rather than licensed operators, a s well as potentially eliminating humans as a layer of defense-in-depth. Separately, remote opera tions would require the NRC staff to reassess current requirements for the application of h uman factors engineering (HFE).

Historically, operators could be expected to take advantage of being co-located with the reactor facility in order to receive sensory feedback (e.g., noise, vib rations, local observation of conditions) that would augment the information otherwise provid ed to them through the plants instrumentation and control (I&C) interfaces. Such information can be very useful in conditions of I&C failures, particularly when highly automated systems are involved. Autonomous and remote operations approaches represent a shift in operator capa bilities that needs to be carefully considered. The NRC staff has been working to develop the needed HFE tools in these

8 As used in this paper, the terms automation and autonomous have distinct meanings. Automation refers to automated processes. The term automated, in turn, is defined here as the independent performance of tasks through the application of technology and absent continuous input from an operator. It should be noted that the proposed 10 CFR Part 53, Risk Informed, Technology-Inclusive Regulatory Frameworks for Commercial Nuclear Plants, rulemaking would define automation as a device or system that accomplishes (partially or fully) a function or task.

9 M. R. Endsley, From here to autonomy: lessons learned from human-automation research, Human factors, vol. 59, no. 1, pp. 5-27, 2017.

10 The regulations at 10 CFR 50.34(f)(2)(iii) require, in part, that applicants provide for NRC review a control room design that meets state-of-the-art human factors engineering principles.

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areas as part of broader efforts at developing an HFE framework for the review of advanced reactor designs.11, 12

Autonomous and Automatic Operations

Tasks may be fully or semi-automated, creating variations in th e required degree of human oversight and control. It is important to note that increased u se of automation is distinct from autonomy. Autonomy is considered the ability to operate with co mplete independence from human control, while automation refers to the machine execution of what were formerly human tasks.13 Thus, autonomous operation may not rely on high levels of auto mation and could be achieved through simplicity (e.g., reliance on inherent safety characteristics and robust passive systems) of an advanced reactor design. 14 The ability of a given design to demonstrate autonomy in its safety performance could also be a significant factor in justifying a remote operational concept for a micro-reactor facility.

The term autonomous operation does not have a commonly accept ed definition in the nuclear industry. For example, a designer may potentially refer to a sy stem as being autonomous or fully automated without the inclusion of any artificial intel ligence while another party might assume that autonomy implies the inclusion of artificial intell igence.15 Autonomous systems can, when the necessary capabilities are provided, potentially respo nd to situations beyond those explicitly programmed or anticipated in the design. Thus, auton omous systems can be capable of a certain amount of self-directed behavior and potentially a ct as a proxy for humans in decision-making situations. 16 The incorporation of artificial intelligence will present new considerations for both developers and the NRC staff because it will introduce the potential for automation to take actions other than those that were originall y assumed during the design and licensing processes. Furthermore, the acceptability of allowing AI-driven automation to control safety-significant operations (e.g., those needed to mitigate a ccidents) represents a currently unresolved matter and an area of ongoing discussion among desig ners, researchers, and the NRC staff alike.

Remote Operations and Remote Monitoring

One currently envisioned use for micro-reactors is electrical p ower generation on micro-grids, including instances in which the micro-reactor is the primary s ource of power, as well as those

11 See, for example, SECY-20-0093.

12 As automation is inherently referenced to those tasks historically performed by humans, technological progress tends to gradually influence what may or may not be considered to represent advances in automation. Refer to R. Parasuraman and V. Riley, Humans and automation: Use, misuse, disuse, abuse, Human Factors, vol. 39, no. 2, pp. 230-253, 1997.

13 NUREG-2261, Artificial Intelligence Strategic Plan Fiscal Years 2023-2027, issued May 2023 (ML23132A305), includes a notional framework for artificial intelligence and autonomy levels in commercial nuclear activities. However, NUREG-2261 generally discusses autonomy within the context of artificial intelligence (e.g., machine learning). For that reason, this enclosure relies on the definition of autonomous presented in the discussion of autonomous systems earlier in this section of the enclosure.

14 The concept of autonomy (particularly where safety performance is concerned) potentially not being achieved by complex automation but rather through simple, robust, and highly reliable safety features and characteristics runs counter to commonly used automation hierarchies that tend to instead represent autonomous operation as consisting of a very high degree of automation.

15 Gaining clarity on industry deployment concepts for autonomous operation will be an area of focus under the Next Steps portion of this section.

16 National Academies of Sciences, Engineering, and Medicine. 2022. Human-AI Teaming: State-of-the-Art and Research Needs, The National Academies Press, Washington, DC, 2022.

https://doi.org/10.17226/26355.

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in which it supplies the grid in parallel with other generation assets. Such uses would benefit substantially from the ability to let grid demand change plant output without a human operator serving as an intermediary, such as by permitting grid control centers staffed by non-NRC-licensed individuals to directly control the electrical generat ion of the micro-reactor facility.

However, operations in which a non-licensed individual directly modifies the power level of a nuclear reactor would be precluded by current NRC regulations i n 10 CFR Part 55, Operators Licenses. Thus, proposals for load-following operations will n eed to be considered by the staff on a case-by-case basis to ensure that the approach used will b e consistent with both statute and regulation. As background, AEA section 11r. defines operato rs as individuals who manipulate the controls of utilization or production facilities. The AEA then mandates under section 107 that individuals who operate controls must be lic ensed by the NRC. Notably, the AEA does not define what those controls consist of, thus affo rding the NRC the discretion to establish that definition by regulation. Both 10 CFR 50.2 and 5 5.4 define these controls (when used within the context of nuclear reactors) as consisting of a pparatuses and mechanisms that directly affect the reactivity or power level of the reactor wh en manipulated.

The NRC staff anticipates that mic ro-reactor applicants will propose to operate (or in the case of autonomous reactors, monitor) one or more micro-reactor units f rom a remote location. Such cases raise the question of what communication and safety featu res would be necessary to provide for the reliable and secure monitoring and control of o ne or more micro-reactor units from a remote location. For example, commercial large light-wat er reactor licensees have historically credited human actions for performing certain time -critical operations to meet accident analysis assumptions. Any suitable approach to remote operations would, logically, need to either provide a very high degree of assurance in the a bility of operators to remotely accomplish such actions or, alternatively, eliminate reliance o n such actions for the achievement of safety functions. The absence of operators on site may poten tially remove any opportunity for local, backup actions should remote operations be unsuccessful for any reason.

The practical implementation of remote operation of commercial nuclear power plants has not yet been explored extensively, so there is a paucity of operati ng experience to draw from to inform future approaches to remote operations. A study of the f easibility of an unattended light-water reactor design in the early 1960s determined that the con cept depended on whether safety systems could be designed to a level of reliability high enough to preclude the need for regular maintenance, thereby necessitating a relatively simple design.17 At present, areas of regulatory guidance are needed to further develop the concept. Centrally, these include resolving issues related to the design attributes that would be necessary to support a safety determination for a remotely operated reactor facility. For exa mple, the potential for loss of remote-control capability may warrant requiring remotely operat ed reactors to meet criteria comparable to those proposed for self-reliant-mitigation facil ities under 10 CFR Part 53 based on a need to robustly demonstrate safety in the absence of any opportunity for human intervention.

Instrumentation and Control

As part of the plants safety analysis, a systematic, comprehen sive assessment of potential internal and external hazards, including the potential for huma n-induced events, and their consequences must be performed. Such an assessment should inclu de any potential hazards stemming from the use of I&C systems within the context of auto nomous or remote operations.

17 M.W. Rosenthal, et al., The Feasibility of an Unattended Nuclear Power Plant, Oak Ridge National Laboratory Report, ORNL-2985, August 1960 (https://www.osti.gov/servlets/purl/4134966).

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These potential hazards must be identified, analyzed, and appro priately addressed (e.g.,

prevented or mitigated) as part of the I&C design for safety. W hen I&C systems are relied on to satisfy the overall nuclear power plant performance objectives, the design of the risk significant I&C systems must meet applicable regulations for safety. The I& C design, used for functions such as sensing, controlling, displaying, and monitoring for th e plant, should be sufficiently reliable and robust commensurate with its safety significance a s required by the regulations.

The applicable regulations for I&C, such as those under 10 CFR Part 50 or 10 CFR Part 52, are generally technology inclusive and performance-based as they pr imarily require the adequate demonstration of reliability (e.g., testing, surveillance, fail -safe design, and quality) and robustness (e.g., redundancy, independence, diversity, defense-in-depth, deterministic behavior, and qualification) independent of specific I&C techno logies and provide licensees with flexibility to determine how to meet the established performanc e criteria. Similarly, NRC staff guidance that is risk-informed, performance-based, and technolo gy-inclusive will be used to assess whether the applicant demonstrates how the specified I&C systems support the overall nuclear power plant performance objectives for a particular pla nt design. For example, the NRC staff has developed a Design Review Guide (DRG): Instrumentati on and Controls for Non-Light-Water Reactor (Non-LWR) Reviews, revised February 26, 2021 (ML21011A140),

which provides guidance for the NRC staff to use in reviewing t he I&C portions of applications for advanced non-light-water reactors that follow Regulatory Gu ide (RG) 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Meth odology to Inform the Licensing Basis and Content of Applications for Licenses, Certi fications, and Approvals for Non-Light-Water Reactors, issued June 2020 (ML20091L698), within t he bounds of existing regulations.

Some of the biggest technical challenges faced by micro-reactor developers will be adapting or developing measurement processes to operate in significantly mo re cramped or inaccessible spaces that limit maintenance access. Furthermore, I&C equipmen t must be rugged enough to handle not only the high temperatures and high pressures in som e advanced reactor designs but also the long-term effects of the coolant on the sensor int erface. High coolant temperature or corrosivity will affect the operational lifespan of I&C equi pment and contribute to measurement uncertainty. Micro-reactors that operate with therm al or fast neutron spectra may also have different sensor/actuat or design requirements due to a wide variety of fuels, higher operating temperatures, and flexible operation modes. Additiona l regulatory guidance may be needed to address these potential challenges for I&C equipment.

Cybersecurity

The current power reactor cybersecurity requirements in 10 CFR 73.54, Protection of digital computer and communication syst ems and networks, apply to (1) digital computer and communication systems and networks that are associated with saf ety-related, important-to-safety, security, and emergency preparedness (SSEP) functions a nd (2) support systems and equipment that, if compromised, cou ld adversely impact SSEP functions. To implement the requirements, licensees identify critical digital assets (CDAs) that must be protected against cyberattacks. In 10 CFR 73.54, the NRC does not specifically address autonomous or remote operations; however, the performance-based nature of the regula tion allows for both through application of appropriate cybersecurity considerations in a li censees cybersecurity plan. Under 10 CFR 73.54(a), applicants for an operating license under the provisions of 10 CFR Part 50 and holders of a combined license under the provisions of 10 CF R Part 52 are required to address the security of digital computer and communication syst ems and networks in their cybersecurity plans. Data communication pathways that could adv ersely impact SSEP functions 13

would be within the scope of digital computer and communication systems and networks required to be protected under 10 CFR 73.54(a) and would requir e protection against cyberattacks, up to and including the design basis threat as de scribed in 10 CFR 73.1, Purpose and scope.

Depending on the level of autonomy, micro-reactor designers wou ld likely propose using data connections with wired, wireless, or a combination of both path ways to communicate with critical systems and CDAs. A defensive computer security architecture th at employs any type of remote access could be vulnerable to cyberattacks, such as unauthorize d remote access, complete denial of service, or denial of authorized remote access functi ons. The level of autonomy, remote operations, and remote monitoring are important aspects in understanding the associated cybersecurity risks. A defensive computer security a rchitecture that incorporates remote operation technology needs to ensure the confidentiality, integrity, and availability of digital computer and communication systems and networks associa ted with SSEP functions.

Additionally, this architecture w ould have to be implemented as part of a mutually supportive framework that includes broader physical protection considerati ons, such as physical security and access authorization.

Stakeholder Perspectives

As noted earlier in this section, the NRC staff has engaged in many stakeholder and pre-application meetings with micro-reactor developers. In these in teractions, the NRC staff has observed high interest in the incorporation of autonomous and r emote operational characteristics into plant designs. These operational concepts have included, in part, remote monitoring of facilities, automatic load-following, remote loca tion of the shift technical advisor, automated reactivity control, elimination of licensed operator staffing outside of startup conditions, and related approaches.

The Nuclear Energy Institute (NEI) anticipates that many micro-reactors will include features related to automatic operations, remote operation, and remote m onitoring.18 NEI has also recognized that micro-reactor designs submitted to the NRC will need to include detailed information on both autonomous and remote operation features. P art of the NEIs observations reflect the perspective that a traditional control room might n ot be necessary on account of technological advances. Notably, the NEI went so far as to spec ulate that some micro-reactors could be fully autonomous and not require operators outside of activities such as initial startup, testing, and the movement of fuel. More recently, NEI has expre ssed the perspective that near-term deployment models will include systems that allow automati c response by the reactor to demand from the grid but that such systems are unlikely to incorporate artificial intelligence.

Sandia National Laboratories (SNL) has observed that clear defi nitions do not yet exist for automation usage, even though high degrees of automation are be ing factored into micro-reactor designs. 19 Additionally, SNL observed gaps in the understanding of how th e humans associated with micro-reactor operations will be involved in bo th onsite and remote tasks and decision-making. A key point noted by SNL was the importance of specifying both the levels of automation and the degrees of human engagement across micro-rea ctor systems.

18 Nuclear Energy Institute, Micro-Reactor Regulatory Issues, NEI White Paper, November 2019 (ML19319C497).

19 Sandia National Laboratories, Human Factors Considerations for Automating Microreactors, Sandia Report SAND2020-5635, June 2020 (ML20175A117).

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During pre-application meetings with the NRC staff, micro-react or developers have proposed to use some degree of remote control and monitoring associated wit h safety and security functions. In remote operation and monitoring, a reliable data communication pathway to CDAs associated with SSEP functions is essential to securely transmit operational information on the reactor.

Near Term Strategy

To support the NRCs HFE reviews of advanced reactor license ap plications under 10 CFR Part 53, the NRC staff recently completed development of draft DRO-ISG-2023-03, Development of Scalable Human Factors Engineering Review Plans (ML22266A072). The draft guidance describes a method for scaling the scope and dep th of HFE reviews for non-light-water reactor technologies such as micro-reactors, enabling the NRC staff to readily adjust the focus and level of its HFE review efforts considering factors s uch as risk insights and the unique characteristics of the design or facility operation (e.g., remo te or autonomous operation).

Although the guidance was developed to support reviews of appli cations submitted under the proposed 10 CFR Part 53, the NRC staff is considering the poten tial use of the general methods described in this guidance to scale HFE reviews of micro-reacto r applications submitted under 10 CFR Part 50 or 10 CFR Part 52.

For near-term license applications that include proposals for r emote or autonomous operation, the NRC staff would use available guidance 20 to assess compliance with the applicable 10 CFR Part 50 or 10 CFR Part 52 regulations and applicant requests fo r exemptions, as needed. The following are examples of areas where deployment models that in clude remote or autonomous operations may require exemptions from existing requirements or raise significant questions of interpretation not addressed under existing guidance.

Remote Operations

Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 contains the general design criteria (GDC), which establish the minimum requ irements for the principal design criteria (PDC) for water-cooled nuclear power plants si milar in design and location to plants for which construction permits have been issued by the C ommission. Appendix A also establishes that the GDC are considered to be generally applica ble to other types of nuclear power units and are intended to provide guidance in determining the PDC for such other units.

GDC 19, Control room, requires (for water-cooled reactors), i n part, that a control room be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under acciden t conditions, including loss-of-coolant accidents. RG 1.232, Revision 0, Guidance for Developi ng Principal Design Criteria for Non-Light-Water Reactors, issued April 2018 (ML17325A611), des cribes the NRCs guidance on how the GDC may be adapted for non-light-water reactor desig ns.

As noted above, 10 CFR 50.34(f)(2)(iii) requires, in part, a co ntrol room design that reflects state-of-the-art human factor principles. For applications that include proposals for remote operation of a nuclear reactor, the NRC staff intends to apply this requirement to any facility from which a nuclear reactor can be operated remotely. In addit ion, the NRC staff would need to determine whether providing a facility from which the reactor c an be operated remotely would

20 As described under the section for Next Steps, the staff is researching gaps in the NRCs human factors engineering regulatory framework that may relate to remote and autonomous operations and develop the technical bases for any new guidance that may be needed.

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obviate the need for an onsite control capability beyond that w hich would otherwise be provided by equipment at appropriate locations outside the control room with (1) a design capability for prompt hot shutdown of the reactor, including necessary I&C to maintain the unit in a safe condition during hot shutdown, and (2) a potential capability f or subsequent cold shutdown of the reactor through the use of suitable procedures.

The requirements in 10 CFR 50.54(m) specify minimum licensed op erator staffing. The requirements are stated largely in terms of operators being req uired to be present at the facility or on site. As written, the requirements do not address a mod el in which operation of a nuclear reactor would be performed from a location other than on site. For applications that include proposals for remote operation of a nuclear reactor, the applic ant would need to request an exemption from requirements in 10 CFR 50.54(m), unless the depl oyment model included maintaining onsite licensed operator staffing that meets the cu rrent staffing requirements.21 In July 2005, the NRC staff published NUREG-1791, Guidance for As sessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffin g Requirements Specified in 10 CFR 50.54(m) (ML052080125).22 The NRC staff developed the guidance to support reviews of staffing for advanced reactors in which the concept of opera tions differed from that of the large light-water reactors that are the basis for the current s taffing regulations. The NRC staff intends to evaluate staffing exemption requests associated with applications proposing remote operations using NUREG-1791 as the NUREG addresses not only pot ential reductions in staffing but also the possibility of operation from remote faci lities and portable devices.

However, as remote operations concepts mature, the NRC staff in tends to continue to evaluate the need to supplement existing guidance and develop new guidan ce.

The NRCs requirements in 10 CFR Part 26, Fitness for Duty Pro grams, apply to, among others, licensed operators at light-water reactor nuclear power plants currently licensed under 10 CFR Part 50 or 10 CFR Part 52. In developing these requireme nts, the NRC did not consider the possibility of remote operations where the reactor may be c ontrolled by individuals who are not at the reactor site. Therefore, the NRC staff needs to asse ss how 10 CFR Part 26 would apply to license applications proposing remote operation of a n uclear reactor facility and address any issues that are identified through an appropriate r egulatory process.

Autonomous Operations

The requirements in 10 CFR 50.54(i), (j), (k), and (m) address onsite operator staffing and having a licensed operator at the controls during facility oper ation. They require that only licensed operators may manipulate controls and that apparatus a nd mechanisms, other than controls, that may affect reactivity or power level be manipula ted only with the knowledge and consent of an operator or senior operator. License applications submitted under 10 CFR Part 50 or 10 CFR Part 52 for autonomous operations of nuclear reactors that, for example, would not include licensed operator staffing in a control room would need to include requests for

21 Staffing operators both on site and at a remote operations facility would not be an economical long-term strategy but is an option that licensees might exercise if it is deemed a viable pathway toward developing a performance-based rationale for requesting an exemption from onsite staffing requirements at some point following initial plant startup.

22 More recently the NRC staff developed draft interim staff guidance to augment NUREG-1791 to support NRC staff review of advanced reactor staffing plans under 10 CFR Part 53 (i.e., DRO-ISG-2023-02, ISG Augmenting NUREG-1791, Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m), for Licensing Plants under Part 53 (ML22266A068)).

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exemptions from the pertinent requi rements. Also, an applicatio n that would permit a non-licensed grid operator to change the plants output in load-fol lowing operations may need to include requests for exemptions from these requirements, where relevant. Beyond this, additional regulatory implications will also need to be address ed, such as requirements for a control room and a remote shutdown capability. The NRC does not currently have guidance specific to the evaluation of such requests. Additionally, as n oted previously, the AEA places certain mandates upon the NRC to license those individuals who will operate the controls of utilization facilities. However, as described in the following subsection, Next Steps, the NRC staff plans to further develop its understanding of deployment models for factory-fabricated micro-reactors, identify gaps in the existing guidance, and dev elop the technical bases for any new guidance that may be needed.

Furthermore, it should be noted that the Advisory Committee on Reactor Safeguards has stated its opposition to the notion of unsupervised operations, irresp ective of the degree of automation involved, and included a recommendation to this effect in its l etter dated November 22, 2022, to Chair Hanson (ML22319A104). The potential for a future reactor design to demonstrate adequate safety performance while operating autonomously may re present an area in which the Commission will need to make a policy decision regarding whethe r some form of oversight by a human operator will be required even when rendered unnecessary for safety or defense-in-depth (e.g., as a conservative measure for public confidence).

Cybersecurity

Control commands needed for remote operation require bidirectio nal data communication.

Bidirectional data flow would create digital architectures that vary from those used by plants in the current operating fleet, which only allow unidirectional da ta flow from high levels of security to lower levels. Fully understanding the architecture will be k ey in providing insights about the attack surface and potential attack vectors. It is likely that protection against disruption or malicious control for micro-reactors will rely heavily on prope rly implemented defensive computer security architectures. Furthermore, the level of auto nomy and the capabilities of the autonomous technologies that would be used to replace humans fo r remote operations would be equally important to understand.

Next Steps

The NRC staff plans to further develop its understanding of the industry deployment models for factory-fabricated micro-reactors with respect to industry plan s for remote and autonomous operations, identify any gaps in the existing HFE regulatory fr amework needed to address the deployment models, and develop the technical bases for any new guidance that may be needed. Examples of specific areas for further staff considerat ion to inform this effort currently include the remote monitoring of operations, remote control of plant functions, physical distance of remote operations facilities from reactor facilities, means of communicating monitoring and control signals, relative degrees of reliance on automation, le vels of automation employed, potential use of artificial intelligence in control functions, and the related implications associated with utilizing passive features, inherent characteristics, or m anual actions to achieve safety functions.

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4. Transportation of Fueled Factory-Fabricated Modules

Deployment Model Considerations for Transportation

Factory-fabricated micro-reactor developers (and potentially de velopers of floating nuclear power plants that use reactors with higher power levels) envisi on transporting fueled factory-fabricated modules from a factory to the deployment site for op eration and later removing fueled modules from the deployment site at the end of operation. Featu res to preclude criticality, as proposed by the NRC staff in this paper, would be needed so tha t the factory-fabricated module would not be considered to be in operation when loaded with f uel.

Shipment of a factory-fabricated module containing fuel would b e subject to the radioactive material transportation requirements in 10 CFR Part 71, Packag ing and Transportation of Radioactive Material, and in Department of Transportation (DOT ) regulations in Title 49 of the Code of Federal Regulations (49 CFR). Shipment would also be subject to the security and possession requirements applicable to the licenses held by the fabricator for possession of the modules, special nuclear material, and any other radioactive ma terial contained in the modules.

Also, as stated in this paper, the NRC staff assumes that the f abricator will obtain a manufacturing license under 10 CFR Part 52 for the factory-fabr icated modules and that the manufacturing license would authorize possession of the modules at the factory.

The requirements in 10 CFR 70.20a, General license to possess special nuclear material for transport, and 10 CFR 70.20b, General license for carriers of transient shipments of formula quantities of strategic special nuclear material, special nucle ar material of moderate strategic significance, special nuclear material of low strategic signifi cance, and irradiated reactor fuel, contain provisions for general licensees to possess special nuc lear material and irradiated fuel during transport and storage incident to transport. The general license requirements in 10 CFR 70.20a and 10 CFR 70.20b also contain physical protectio n requirements to follow the appropriate sections in 10 CFR Part 73, Physical Protection of Plants and Materials, for material transport and storage incident to transport. However, the regulations for physical protection of special nuclear material in transport do not expl icitly include requirements for special nuclear material that is loaded in a utilization facili ty during transport. As discussed in the main body of this paper, the NRC staff will consider whether additional Commission engagement is needed related to physical security requirements for factory-fabricated micro-reactors, including for transportation.

The manufacturing license application must also address certain aspects of shipment of a reactor module. The regulations at 10 CFR 52.157(f)(26)(iv) req uire that the application for a manufacturing license include a description of the proposed pro cedures for shipment of the reactor module. These procedures (which may differ from the ope rating procedures submitted for a package approval under 10 CFR Part 71) govern preparation of the reactor for shipping, performing the shipment, and verification of the reactors cond ition upon receipt at the site to minimize the potential that the reactor arrives at its destinat ion and is unable to operate within the parameters of the license for the deployment site.

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Packaging and Transportation

The regulations in 10 CFR Part 71 contain requirements for Type B,23 Type B fissile (Type BF),

and Type A fissile (Type AF) material packages and their transp ortation. The regulations that apply to the package will depend on the contents. Transportatio n packages for factory-fabricated modules approved under 10 CFR Part 71 may consist of the module itself or the module plus an additional overpack or other materials, as neede d, to meet the packaging requirements. Packages for transporting a module from the facto ry to the deployment site could be either a Type AF or Type BF package, as defined in 10 CFR Pa rt 71. Selection of the appropriate package would depend on the enrichment isotopic con tent of the uranium in the loaded fuel and whether the module was operated at the factory, which would determine whether there is greater than a Type A quantity of radioactive material in the package.

In addition to the requirements in 10 CFR Part 71, the packagin g and transport of licensed material are also subject to other NRC requirements and to the regulations of other agencies. A factory-fabricated module, whether shipped before operation or after operation, would be subject to the requirements in 10 CFR Part 20, Standards for P rotection Against Radiation, and 10 CFR Part 21, Reporting of Defects and Noncompliance, r egardless of the contents of the package or the licensing pathways chosen by the Commission in response to this paper.

While in transit, a license issued pursuant to 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, would be required for possession of the spec ial nuclear material in the fuel loaded in the module, and a byproduct material license issued p ursuant to 10 CFR Part 30, Rules of General Applicability to Domestic Licensing of Byprod uct Material, would be required if the module had been operated for testing and contained fissi on products.

The NRC also co-regulates transportation with the DOT. The DOT regulates shipment of all classes of hazardous material, including radioactive materials. The shipper and carrier are subject to DOT regulations and, depending on the components of the package, may be subject to more than one hazard class under DOT regulations. DOT regula tions authorize shipment of fissile material and radioactive material in some NRC-approved packages under 49 CFR 173.415 and 49 CFR 173.416, respectively. However, if th e shipper seeks an NRC package approval under 10 CFR 71.41(c) or seeks an exemption fr om one or more of the package approval requirements in 10 CFR Part 71, subparts D, E, and F, the shipper must obtain a special permit issued by the DOT as these package appr ovals are not automatically authorized in DOT regulations.

Front End Transport

As discussed below, 10 CFR Part 71 is adequate for approving a factory-fabricated module for shipment. A package used to transport a fueled factory-fabricat ed module must be evaluated against the package performance requirements in 10 CFR Part 71 for the type of package that the module fabricator or licensee proposes to use to ship the m odule. A fueled factory-fabricated module that has not been operated at a factory and c ontains commercial grade uranium enriched to less than 20 weight percent in the uranium-235 isotope would be classified as a Type AF package. However, if the package contains low-enri ched fuel that includes reprocessed or downblended uranium, then the package would like ly be classified as a Type BF

23 Type B packages contain a quantity of radioactive material greater than a Type A quantity. The NRC defines a Type A quantity of material and a fissile material package in 10 CFR 71.4.

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package, depending on the quantity of impurities in the fuel an d the A1 or A2 value24 for the mixture. Transport of a module that has been operated at a fact ory could call for a Type AF or Type BF package depending on the quantity of radionuclides gene rated during operation and the radionuclides initially present (e.g., if the fuel was repr ocessed or downblended). Based on the expected power level and duration of testing and the time b etween the completion of testing and shipment, the applicant should determine the quantity of ra dionuclides that would be present at the intended time of shipment, using Appendix A to 1 0 CFR Part 71, and determine whether the contents constitute a Type A or Type B quantity of material. The regulations in 10 CFR 71.31, Contents of application, require the applicant for a package approval to describe the contents and determine, based on the proposed cont ents, whether the package would be a Type AF or Type BF package.

Both Type AF and Type BF packages would be subject to the tests and conditions for normal conditions of transport and hypothetical accident conditions an d required to maintain criticality safety in accordance with 10 CFR 71.55, General requirements f or fissile material packages, and 10 CFR 71.59, Standards for arrays of fissile material pac kages, and dose rates for normal conditions of transport below the criteria in 10 CFR 71. 47, External radiation standards for all packages. However, there would be no containment or do se rate criteria after hypothetical accident conditions for the Type AF package. For T ype BF packages, the package designer would be required to show that the package meets the h ypothetical accident condition dose rate and containment criteria in 10 CFR 71.51, Additional requirements for Type B packages, in addition to maintaining criticality safety and me eting the dose rate criteria in 10 CFR 71.47.

Shipment of Factory-Fabricated Modules Between NRC-Licensed Sit es or Back End Transport

The point at which a factory-fabricated module would transition from a Type AF package to a Type BF package depends heavily on the module design and the po wer level and duration of operation. The NRC staff estimates that after half a day of ful l power operation or the equivalent, the module would likely contain greater than a Type A quantity of radionuclides. Presuming that the factory-fabricated module is operated at full power for mor e than a single day, the applicable regulatory requirements would very likely be those for a Type B F package.

Near Term Strategy

The regulations in 10 CFR Part 71 contain performance-based req uirements for packaging and transportation of radioactive material. The NRC regulations hav e been harmonized with the International Atomic Energy Agency standards in the 2009 Editio n of TS-R-1, Regulations for the Safe Transport of Radioactive Material, to facilitate inte rnational transport.25 However, under the current regulatory framework, the NRC may approve alt ernate standards for packages that may not meet all of the packaging requirements in 10 CFR Part 71. The NRC staff intends to use the existing regulatory framework in 10 CF R Part 71 to review transportation of commercial fueled factory-fabricated modules in the near ter m. Specifically, 10 CFR 71.41(c) allows for environmental and test conditions different from tho se in 10 CFR 71.71, Normal conditions of transport, and 71.73, Hypothetical accident con ditions, if the shippers controls

24 Appendix A, Determination of A1 and A2, to 10 CFR Part 71 contains the A1 and A2 values for individual radioisotopes and the method for calculating the aggregate A1 and A2 value for a mixture of radionuclides.

25 The NRC is in the process of harmonizing 10 CFR Part 71 with the International Atomic Energy Agency safety standards in Specific Safety Requirements No. 6 (SSR-6) (2018 Edition). See proposed rule dated September 12, 2022, (87 FR 55708) Harmonization of Transportation Safety Requirements with IAEA Standards (RIN 3150-AJ85; NRC-2016-0179).

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provide safety of the shipment equivalent to that provided by m eeting the regulations. Further, 10 CFR 71.41(d) provides for one-time special package authoriza tion if the application demonstrates that compliance with the regulations is impractica ble and the safety standards established by the regulations have been met through alternativ e means. Finally, an applicant may request an exemption as specified in 10 CFR 71.12, Specifi c exemptions. Each of these alternatives has limitations, as described in more detail below.

The requirements in 10 CFR 71.41(c) provide for alternate envir onmental and test conditions for a package that, when subjected to the environmental conditions required by the regulations, in conjunction with one or more of the tests for normal conditions of transport or hypothetical accident conditions, cannot meet the post-test criteria. Use of the alternate test criteria in 10 CFR 71.41(c) has several limitations. An applicant for packa ge approval cannot eliminate the test but rather can reduce the severity of the test (e.g., the applicant can use a 20-foot drop instead of 30-foot drop but cannot substitute a different test) so that the package can meet the post-test criteria. In addition to the alternative environmenta l conditions or test criteria, the applicant must submit additional controls that the shipper can exercise to provide an equivalent level of safety for the shipment. Because the regulations in 10 CFR 71.41(c) do not offer alternate post-test criteria, the applicant would still need to meet the regulatory limits for dose rate, containment, and criticality safety. Differing post-test criteria can be approved only through exemption.

After its experience with issuing the exemptions for the Trojan reactor vessel (see SECY-98-0231, Authorization of the Trojan Reactor Vessel Packa ge for One-Time Shipment for Disposal, dated October 2, 1998), 26 the NRC noticed a need for a provision for a special package authorization for one-time shipment of large components that do not meet the criteria for shipment as low specific activity packages or surface conta minated objects.27 Therefore, the NRC promulgated 10 CFR 71.41(d), which provides for a special p ackage authorization for one-time shipments when compliance with the other provisions of 10 CFR Part 71 is impracticable. To obtain a special package authorization, the a pplicant must demonstrate that the overall level of safety in transport for these shipments is at least equivalent to that which would be provided if all the applicable requirements had been m et. Because 10 CFR 71.41(d) is limited to one-time shipments, a different application and a n NRC review resulting in a different approval would be needed for each shipment for each f actory-fabricated module, likely making this option cost-prohibitive.

If neither 10 CFR 71.41(c) nor 10 CFR 71.41(d) can be used, lic ensees can request an exemption from the regulations pursuant to 10 CFR 71.12. Throug h exemption, licensees can provide alternate environmental conditions and tests and alternate post-test criteria. The exemption request must contain sufficient technical information for the NRC staff to determine that the request is authorized by law and will not endanger lif e or property or the common defense and security. The exemption request should be accompani ed by an environmental report because the NRCs practice is to not apply the categoric al exclusion in 10 CFR 51.22(c)(13) for package designs for packages to be use d for the transportation of licensed materials to packages approved by relying on an exemp tion rather than the specific package approval regulations. 28 In addition, each licensee making a shipment needs to request

26 https://www.nrc.gov/reading-rm/doc-collections/commission/secys/1998/secy1998-231/1998-231scy.pdf.

27 For the definitions of low specific activity and surface-contaminated object, see 49 CFR 173.403, Definitions. For the exemption from most of the requirements in 10 CFR Part 71 for low-specific-activity packages and surface-contaminated objects, see 10 CFR 71.14(b)(3).

28 See the final rule Compatibility With IAEA Transportation Safety Standards (TS-R-1) and Other Transportation Safety Amendments, 69 FR 3698, 3744 (January 26, 2004).

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a separate exemption, because an exemption cannot be made gener ically applicable to multiple licensees. Further, the DOTs regulations do not specifically a uthorize NRC-issued exemptions as a package approval; therefore, each licensee would need a DO T-issued special permit for its shipment.

Next Steps

The NRC staff will continue to engage with factory-fabricated m icro-reactor developers and potential package applicants to discuss their plans for package approval through pre-application activities and through the periodic Advanced Reactor Stakeholde r Meetings. As appropriate, depending on the package approval plans for factory-fabricated modules, the NRC staff will evaluate the need for future Commission papers, rulemaking, and guidance, including for security.

5. Storage of Fuel After Irradiation in a Power Reactor

Deployment Model Considerations

Irradiated fuel for factory-fabricated micro-reactors (and adva nced reactors in general) may not need to be water-cooled immediately after shutdown to ensure sa fety because the irradiated fuel may be able to withstand higher temperatures without failu re than fuel historically used in light-water reactors. As discussed below, however, the irradiat ed fuel must, among other things, be withdrawn from the reactor and have undergone at least 1 yea r of decay since being used as a source of energy in a power reactor before it may be stored i n an independent spent fuel storage installation (ISFSI) regulated under 10 CFR Part 72, L icensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioacti ve Waste, and Reactor-Related Greater than Class C Waste. Therefore, absent exemptions from the scope defined in 10 CFR 72.2(a) and other related requirements in 10 CFR Part 72, irradiated fuel withdrawn from the reactor that has not undergone decay for at least 1 year would be required to be licensed under 10 CFR Part 50 or 10 CFR Part 52, similar to a spent fuel pool for a light-water reactor.

As provided in 10 CFR 72.2(a)(1), ISFSIs licensed under 10 CFR Part 72 are limited to the receipt, transfer, packaging, and possession of [p]ower reacto r spent fuel to be stored in a complex that is designed and constructed specifically for stora ge of power reactor spent fuel aged for at least one year, other radioactive materials associa ted with spent fuel storage, and power reactor-related GTCC [greater than class C] waste in a so lid form in an independent spent fuel storage installation (ISFSI). Similarly, the defini tion of spent fuel in 10 CFR 72.3, Definitions, includes criteria that the fuel has been withdra wn from a nuclear power reactor following irradiation and has undergone at least 1 years decay since being used as a source of energy in a power reactor.29

29 The NRCs definition of spent fuel in 10 CFR 71.4, Definitions, is the same as that currently found in 10 CFR 72.3, Definitions, for ISFSIs. The transportation and packaging requirements in 10 CFR Part 71 do not use the term spent fuel, so the only place where this term exists in the transportation regulations is in the definition of spent fuel in 10 CFR 71.4. Because there are no package approval standards or transportation requirements in 10 CFR Part 71 stating that the fuel must undergo decay for at least a year prior to transport, the NRC can approve package designs for shipment of fuel that has undergone decay for less than 1 year.

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The regulatory history of 10 CFR Part 72 provides insight into the basis for this one-year decay period. In the proposed rulemaking for 10 CFR Part 72 dated Oct ober 6, 1978 (43 FR 46309),

the NRC stated:

The storage of spent fuels under water is only necessary for th ose fuels which have not undergone sufficient aging since their discharge from a reactor to make cooling by some other means feasible.

The proposed rule is applicable only to aged fuel, with more than one years decay since reactor shutdown. Aged spent fuel, having lost the short-lived radionuclides by decay, need not have a high degree of protecti on from weather extremes, tornadoes, or tornado generated missiles.

At the time, water cooling of commercial light-water reactor sp ent fuel was deemed necessary prior to placement in an ISFSI to ensure that the fuel would no t overheat. Storage of spent fuel in an ISFSI was not limited to water pool installations. Furthe r, in the final rulemaking dated November 12, 1980 (45 FR 74694) the NRC stated:

The long-lived radionuclides present in spent fuel are proporti onal to burnup; but within the limits of expected burnups, this is not a significan t factor for spent fuel aged more than one year.

The one-year decay stipulation has been retained as this is a b asis for the requirements of Part 72, i.e., the presumption is made that no short-lived radionuclides are present, and the levels of volatile radioacti ve materials are very substantially reduced.

Inasmuch as the definition of spent fuel eligible for storage i n an ISFSI [Section 72.3] specifies that the fuel must have undergone at least a ye ars decay since its irradiation in a power reactor, any facility for temporary stor age of fuel irradiated in a power reactor which has not undergone a years decay would be licensed under Part 50 rather than Part 72.

As stated in 10 CFR 50.1, Basis, purpose, and procedures appli cable, the purpose of 10 CFR Part 50 is to provide for the licensing of production and util ization facilities, so the reference to licensed under 10 CFR Part 50 in the 1980 final rulemaking sh ould be understood in that context. Since that 1980 rule, the NRC has established 10 CFR P art 52 as an additional available path to license nuclear power reactors.

Near Term Strategy

Dry cask storage designs for use at a generally licensed ISFSI or an ISFSI with a specific license for storage of spent fuel that has undergone decay for at least a year would be approved under 10 CFR Part 72. However, storage of irradiated fuel withd rawn from a reactor that has undergone decay for less than 1 year would be licensed under ei ther 10 CFR Part 50 or 10 CFR Part 52, unless exemptions from the requirements in 10 CFR Part 72 were granted to allow issuance of an ISFSI license.

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Next Steps

The NRC staff intends to engage with stakeholders as they furth er develop their strategies for handling and storage of irradiated and spent fuel generated in factory-fabricated micro-reactors.

6. Decommissioning Process, Decommissioning Funding Assurance, and Refurbishment and Refueling

Deployment Model Considerations

Factory-fabricated micro-reactor deployment models might involv e novel scenarios such as transporting a factory-fabricated module away from the deployme nt site to a facility at a different location for decommissioning at the end of its life or for refu rbishment and refueling before re-deployment. A decommissioning facility might be used to dism antle the factory-fabricated module to recover reusable parts and prepare the waste and spen t fuel for transfer, or it might also include an independent spent fuel storage facility. A refu rbishment and refueling facility might be used to defuel the factory-fabricated module, perform maintenance, refuel the module, and possibly operate the module for testing before re-deploymen t.

In the following discussion, the NRC staff addresses decommissi oning of factory-fabricated micro-reactors away from the deployment site. 30 This would require physically transferring the module from the deployment site to a decommissioning facility a nd legally transferring responsibility for the reactor from the deployment site licensee to the decommissioning facility licensee.31 The decommissioning facility licensee would need the appropria te license(s) to receive the module and its contents. The deployment site licens ee would be responsible for decommissioning the remaining onsite structures and meeting the relevant criteria for license termination and release of the site. The decommissioning facili ty licensee would be responsible for decommissioning the module. Under this scenario, the deploy ment site licensee might be the same entity as the decommissioning facility licensee or there m ight be different licensees. In either case, a factory-fabricated module would need to be cover ed by separate licenses appropriate for the activities to be conducted at the deploymen t site and at the decommissioning facility. If the module were transferred to a refurbishment and refueling facility instead of a decommissioning facility, then the refurbishment and refueling facility licensee would be responsible for the module and its redeployment and the deploym ent site licensee would be responsible for the deployment site decommissioning.

A decommissioning facility or a refurbishment and refueling fac ility might require several NRC licenses depending on the activities to be conducted at the fac ility. A license issued pursuant to 10 CFR Part 30 would be required to receive, possess, and trans fer the byproduct material created by operation of the reactor at the deployment site. A l icense issued pursuant to 10 CFR Part 70 would be needed for receipt, possession, and tra nsfer of the special nuclear material in the form of the irradiated or spent fuel removed fr om the factory-fabricated module and any fresh fuel needed for refueling. A license issued pursu ant to 10 CFR Part 72 would be

30 Factory-fabricated micro-reactors could also be decommissioned at the deployment site following the traditional approach used for large light-water reactors.

31 Transferring legal responsibility for the reactor would require licenses from the NRC (or embedded authorities within existing licenses) because AEA Section 101 requires a license to transfer a utilization facility and a license to receive in interstate commerce, acquire, or possess a utilization facility. The approach contemplated here is different from a license transfer because, instead of transferring control of an existing license from one person to another, the reactor itself would be transferred and become subject to a different license (i.e., the license for the decommissioning facility).

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required if the facility were also to serve as an ISFSI for spe nt fuel, power-reactor-related greater-than-Class-C waste, and other radioactive materials ass ociated with spent fuel storage.

The facility would also need an operating license issued pursua nt to 10 CFR Part 50 or a combined license issued pursuant to 10 CFR Part 52 to receive, possess, and use the reactor in order to operate it (e.g., for testing) after refurbishment and refueling, and for reactor and facility decommissioning. Also, a utilization facility license would be required to receive and possess the factory-fabricated module at a decommissioning facility or a refurbishment and refueling facility even if it is not operated.

At the end of the life of a module or its fuel cycle at the dep loyment site, the deployment site licensee would remove it from service and install features to p reclude criticality that would render the module not in operation as proposed by the NRC staff in this paper. For transportation to either a decommissioning facility or a refurb ishment and refueling facility, the deployment site licensee would be required to prepare the modul e for transport in accordance with the requirements in 10 CFR Part 71, the package approval, and other applicable regulations. Among other requirements, the transportation packa ge would have to meet the criticality safety requirements in 10 CFR 71.55 and 10 CFR 71.5 9, which could be satisfied through installation of the features to preclude criticality. O nce the factory-fabricated module is removed from the deployment site and transferred to the decommi ssioning facility or refurbishment and refueling facility, the module would no longe r be the responsibility of the deployment site licensee. The depl oyment site licensee would be responsible for completing decommissioning of the deployment site consistent with applicab le regulations of 10 CFR Part 20, Subpart E, Radiological Criteria for License Terminat ion, as guided by NUREG-1757, Consolidated Decommissioning Guidance. The deployment site li cense for that reactor could be terminated upon meeting the decommissioning requirements for the remainder of the deployment site structures and systems unique to that reactor. If shared structures, systems, and components were to be used for a replacement factory-fueled module, the NRC would employ a licensing approach that would subject those structures, systems, and components to decommissioning at the appropriate time. To the extent that a d eployment model does not reflect the decommissioning scenarios contemplated in 10 CFR 50.82(a) or 10 CFR 52.110, exemptions or license conditions may be used to ensure that mic ro-reactors and associated structures, systems, and components are decommissioned consiste nt with the underlying purposes of these regulations.

The deployment site licensee would need to establish decommissi oning funding assurance that reflects the cost of removing the module from the site and deco mmissioning it elsewhere in addition to the cost of decommissioning onsite structures in or der to permanently terminate the license and meet the requirements in 10 CFR Part 20, Subpart E. As required by 10 CFR 50.33(k)(1), applicants for a reactor operating license or a combined license must provide a report as described in § 50.75, indicating how reaso nable assurance will be provided that funds will be available to decommission the facility. For power reactor licensees, reasonable assurance consists of a series of steps as provided in paragraphs (b), (c), (e), and (f) of [§ 50.75]. The regulations at 10 CFR 50.75(c) establish the minimum amounts of funding required to demonstrate reasonable assurance of funds for decom missioning by reactor type and thermal power level for pressurized water reactors and boil ing water reactors. However, most current designs for factory-fabricated micro-reactors use non-light-water reactor technology that may involve significantly different decommissio ning considerations and strategies compared to pressurized or boiling water reactors. F urther, 10 CFR 50.75(c) requires that a power level of at least 1,200 megawatts thermal be used in calculating the minimum amounts, which is roughly 500 to 50 times the power level of fa ctory-fabricated micro-reactor designs. The formulas result in required funding assurance in e xcess of $200 million in 2023 25

dollars (in excess of $75 million in January 1986 dollars as sp ecified by 10 CFR 50.75(c)) for each reactor, which, according to industry stakeholders, is not compatible with deployment models. Reliance on use of the minimum formula amount for decom missioning during operations as reflected in 10 CFR 50.75(c) may need to be revis ited as discussed in Near Term Strategy below.

To ensure adequate funding for all activities, a decommissionin g cost estimate would need to consider and account for all activities and waste disposal cost s associated with decommissioning the deployment site and decommissioning the fac tory-fabricated module at a decommissioning facility. It is possible that the decommissioni ng cost estimate for a module could be a predetermined estimate provided by a decommissioning facility licensee or a refurbishment and refueling facility licensee authorized to acq uire such a module (which could be the original fabricator of the module). It is also possible that the deployment site licensee would be the decommissioning facility licensee and a cost estim ate would account for the various component costs to dismantle and dispose of the module and store or transfer the spent fuel. In either case, a preliminary decommissioning plan submit ted with the deployment site license application would need to describe how the decommission ing funds would be accounted for between decommissioning the factory-fabricated module and t he deployment site decommissioning activities. Later, decommissioning plans would be required to be submitted in the form of a post-shutdown decommissioning activities report, a site-specific decommissioning cost estimate, a license termination plan, and a final status s urvey in accordance with 10 CFR 50.82 or 10 CFR 52.110, both titled Termination of lice nse.

Near Term Strategy

The NRC staff intends to use the existing regulatory framework to review applications for licenses related to decommissioning and refurbishment and refue ling activities for factory-fabricated micro-reactors. If the Commission approves the NRC s taffs proposed determination that fueled reactors with features to preclude criticality are not in operation, then the NRC staff will consider this in its review of applications for decommissi oning and refurbishment and refueling facilities. The NRC staff intends to also consider th e approaches for licensing multi-module sites in SECY-11-0079 in relation to decommissioning and refurbishment and refueling facilities that may require 10 CFR Part 50 or 10 CFR Part 52 li censes for each module brought to the facility for decommissioning or refurbishment and refuel ing, especially if refueled reactors will be operated for testing before redeployment.

With respect to decommissioning funding assurance, the NRC staf f would consider site-specific decommissioning cost estimates and requests for exemption from 10 CFR 50.75(c) in the review of applications for operating licenses and combined lice nses for factory-fabricated micro-reactors.32 The proposed draft regulations in 10 CFR Part 53 include the u se of site-specific decommissioning cost estimates.

32 The NRC staff described application of this approach to small modular nuclear reactors in SECY-11-0181, Decommissioning Funding Assurance for Small Modular Nuclear Reactors, dated December 22, 2011 (ML112620358).

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Next Steps

The NRC staff will continue to engage stakeholders on considera tions related to decommissioning and refurbishment and refueling of factory-fabr icated micro-reactors to better understand the range of options under consideration. If the NRC staff identifies issues that involve policy decisions or potential rulemaking, the NRC staff will seek Commission direction through an additional options paper.

7. Siting in Densely Populated Areas

The NRC has a longstanding policy of siting nuclear power react ors away from densely populated centers and preferring areas of low population densit y for reactor sites. As discussed in SECY-20-0045, Population-Related Siting Considerations for Advanced Reactors, dated May 8, 2020 (ML19143A194), the attributes of advanced reactors, including micro-reactors, are expected to reduce the likelihood of accidents and to result in a smaller and slower release of radioactive material in the unlikely event of an accident. Thes e attributes of advanced reactors, if demonstrated, may support siting them closer to population c enters than large light-water reactors typically have been. As such, in SRM-SECY-20-0045, Staff Requirements SECY-20-0045Population-Related Siting Considerations for Advan ced Reactors, dated July 13, 2022 (ML22194A885), the Co mmission approved the NRC staffs proposal to revise the population-related siting guidance in RG 4.7, General Site Sui tability Criteria for Nuclear Power Stations, Revision 3, issued March 2014 (ML12188A053), to prov ide technology-inclusive, risk-informed, and performance-based criteria to assess certain population-related issues in siting advanced reactors.

The NRC staff is updating RG 4.7 to include the alternative pop ulation-related criteria approved by the Commission in SRM-SECY-20-0045. The NRC staff intends th at the guidance will state that, instead of locating a reactor in an area where the popula tion density does not exceed 500 persons per square mile out to 20 miles from the reactor, a n applicant can demonstrate compliance with 10 CFR 100.21(h) by siting a nuclear reactor in a location where the population density does not exceed 500 persons per square mile out to a di stance equal to twice the distance at which a hypothetical individual could receive a cal culated total effective dose equivalent (TEDE) of one rem over a period of one month from th e release of radionuclides following postulated accidents.

While the NRC is revising its population-related siting guidanc e to include alternate means of complying with 10 CFR 100.21(h), the regulations in 10 CFR 100. 21, Non-seismic siting criteria, remain unchanged. This includes the provision in 10 CFR 100.21(b), which requires that [t]he population center distance, as defined in § 100.3, must be at least one and one-third times the distance from the reactor to the outer boundary of th e low population zone. In 10 CFR 100.3, population center distance is defined as the d istance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents.

Further, as discussed in RG 4.7, a densely populated center is considered to contain more than about 25,000 residents and the boundary of the population cente r is determined based on population distribution rather than political boundaries. As su ch, current Commission policy and regulations would preclude siting a commercial power reactor, n o matter the size or type of reactor, within a densely populated center. The allowable dista nce from the reactor to a densely populated center of approximately 25,000 residents would be no closer than 1.33 times the radius of the low population zone (LPZ). In accordance with 10 CFR 50.34(a)(1)(ii)(D)(2),

10 CFR 52.17(a)(1)(ix)(B), and 10 CFR 52.79(a)(1)(vi)(B), the L PZ is required to be of such a 27

size that an individual located on its outer boundary during th e course of the postulated fission product release would not receive a radiation dose in excess of 25 rem TEDE. The size of the LPZ depends on atmospheric dispersion characteristics and popul ation characteristics of the site, as well as aspects of plant design. The NRC staff notes t hat 10 CFR 100.21 is not applicable to research and test reactors. However, testing reac tors are subject to 10 CFR 100.11(a)(3), which requires a population center distanc e of at least one and one-third times the distance from the reactor to the outer boundary of th e LPZ. There are research reactors currently sited within population centers with more th an 25,000 residents, and the NRC staff anticipates license applications for additional research reactors to be sited in densely populated areas.

Deployment Model Considerations

Some micro-reactor license applicants may seek to site reactors at locations that would not conform to the current Commission policy and the regulations in 10 CFR 100.21(b). Such deployment scenarios are being considered for several reasons i ncluding replacing existing coal plants or providing process heat for heating or industrial appl ications, or to provide power to remote communities or smaller grids with relatively small but c oncentrated populations that would be close to a reactor site.

Near Term Strategy

The NRC staff will continue to revise RG 4.7 and will review li cense applications in accordance with current Commission policy that allows alternative populati on-related siting criteria but precludes siting a commercial power reactor, no matter the size or type of reactor, within a populated center of 25,000 residents or more.

Next Steps

The NRC staff will continue to engage with reactor developers a nd prospective license applicants as it revises the guidance in RG 4.7. The NRC staff will inform the Commission if it becomes aware of any license applicants who intend to seek exem ption from 10 CFR 100.21(b) and will raise associated policy issues to the Commission accor dingly.

8. Commercial Maritime Applications

Deployment Model Considerations

The NRC staff is aware of growing interest in commercial mariti me applications of micro-reactors and other reactor technologies for stationary power pr oduction, marine vessel propulsion, production of decarbonized fuels, and other uses. S tationary reactors might be located in ports or other coastal locations or further out from the shore in domestic waters.

Reactors used for commercial maritime vessel propulsion might b e operated solely within U.S waters or internationally, especially in the shipping industry. Reactors used for decarbonized fuel production or other chemical processing, or industrial app lications would typically be stationary power reactors in that they would be moored or ancho red at a fixed site while in operation, but they might be moved among several locations duri ng their lifetime.

The various maritime applications envisioned by developers give rise to many potential legal, regulatory, and policy issues. For example, reactors located in coastal waters may have different environmental considerations than land-based reactors. Such environmental 28

considerations could potentially involve the Coastal Zone Manag ement Act, Magnuson-Stevens Fishery Conservation and Management Act, and Marine Mammal Prot ection Act. In addition, siting in the marine environment may require different approach es to analyses of external hazards, dose modeling, and other matters. Deployment models mi ght also include scenarios in which larger advanced reactors or light-water reactors are fabr icated and potentially fueled in a factory before being transported to the maritime deployment sit e, which could give rise to considerations beyond those examined for micro-reactors in this paper. Deployment models involving nuclear propulsion of commercial maritime vessels, in cluding those where a U.S. flag vessel would operate in international or foreign waters or a fo reign flag vessel would operate in domestic waters, may present regulatory and policy issues on ma tters such as regulatory jurisdiction and domestic and international licensing.

Near Term Strategy

The NRC staff plans to assess the existing regulatory framework and its applicability to the licensing of stationary floating nuc lear power plants, which mi ght use factory-fabricated micro-reactor designs or larger advanced reactor or light-water react or designs.

Next Steps

The NRC staff will monitor developments related to commercial m aritime applications and assess the need for future Commission direction and coordinatio n with other Federal agencies related to deployment of commercial maritime reactors. The NRC staff will also continue to communicate periodically with U.S. Department of Energy (DOE) s taff on maritime reactor activities through the DOEs Maritime Nuclear Application Group.

9. Commercial Space Applications

Deployment Model Considerations

The NRC staff is aware that developers are considering space ap plications of factory-fabricated micro-reactors. Government agencies such as the National Aerona utics and Space Administration and the DOE are encouraging development of the t echnology, primarily for Government projects. The NRC staff is not aware of any fully co mmercial ventures that plan to use micro-reactors for space applications, whether for power ge neration for space vehicles, extraterrestrial installations, or propulsion systems.

In the case of a fully commercial space application of a factor y-fabricated micro-reactor, the NRCs established regulatory jurisdiction and licensing authori ty, subject to the authorities of other agencies such as the DOT, would cover the related terrest rial activities before launch activities.33 Upon initiation of launch activities, the Federal Aviation Adm inistration (FAA) Office of Commercial Space Transportation (a part of the DOT) would ha ve authority.34 Terrestrial activities under the NRCs regulatory jurisdiction would includ e licensing and oversight of the manufacture, construction, potential operation (e.g., for testi ng or technology demonstration,

33 Letter from Samuel J. Collins to Robert DAusilio, dated May 6,1998 (ML20013J130).

34 The Interagency Nuclear Safety Review Board issued a trial use playbook titled, Non-binding Guidance for INSRB and Its Counterparts, dated January 20, 2023, which includes Appendix E: Defining a U.S.

Government Launch versus a Commercial Launch and DOT Authority. The appendix discusses the responsibilities and authorities of FAAs Office of Commercial Space Transportation as they relate to launches of space nuclear systems and states, the definition of what is a commercial launch, from the DOT/FAA licensing perspective, is whether or not the launch or reentry event is commercially conducted.

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transportation, and storage of utilization facilities intended to be deployed at extraterrestrial locations, including space). The NRCs authority under the AEA and 10 CFR Part 50 and 10 CFR Part 52 for domestic licensing and regulation of utiliza tion facilities does not extend to the operation of reactors outside the United States.

Near Term Strategy

If developers engage the NRC staff on terrestrial activities re lated to commercial space applications of micro-reactors, the NRC staff intends to apply the established regulatory framework, as informed by this paper and any resultant Commissi on direction for factory-fabricated micro-reactors.

Next Steps

The NRC staff will monitor developments related to commercial s pace applications, including those involving Government and commercial partnerships, and ass ess the need for future Commission direction.

The NRC staff will continue to engage with other Federal Govern ment agencies on matters of regulatory jurisdiction, licensing, and safety of launches and applications of space nuclear systems, as appropriate, through ongoing interagency activities and the Interagency Nuclear Safety Review Board.

10. Commercial Mobile Micro-Reactors

Deployment Model Considerations

The NRC staff uses the term mobile micro-reactor to refer to a micro-reactor that is intended to be operated at more than one location on an as-needed, where-ne eded basis without a preapproved specific site license. The U.S. Department of Defen se (DOD) Strategic Capabilities Office is working with the DOE to develop and test a demonstrat ion unit for a mobile micro-reactor for military applications as part of Project Pele. NRC licensing is not required for this reactor, but DOD is seeking NRC approval of a transportation pa ckage under 10 CFR Part 71.35 The DOD engaged with multiple vendors who developed proposed de signs for this program.

Some micro-reactor developers have indicated that they may even tually deploy mobile commercial factory-fabricated micro-reactors in this way, such as for disaster relief applications.

However, this deployment model does not appear to be a near-ter m focus for developers.

The NRC has historically issued licenses for land-based reactor s at fixed sites. Issuing separate licenses for each site as the need arises would not support the rapid deployment needed for disaster relief because of the time needed for the licensing pr ocess (safety and environmental reviews, hearings, etc.). To support such rapid deployments, th e NRC would need to issue a license approving potential sites ahead of time (e.g., a licens e that would address safety and environmental issues for all potential operating sites within t he United States or for a set of sites within a region). However, it would be difficult under the curr ent regulatory framework to license commercial mobile micro-reactors for all potential operating si tes. The NRCs approach to safety and environmental reviews presumes that a reactor will operate at a single site. The following

35 The staff is preparing a SECY paper to inform the Commission about a risk-informed methodology that will be used to develop an application for a package approval, utilizing multiple exemptions from Part 71, for a limited number of shipments.

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are two of the specific technical requirements that would need to be satisfied in advance of deployment of a mobile micro-reactor:

(1) The regulations in 10 CFR Part 100, Reactor Site Criteria, establish approval requirements for proposed sites for power and testing reactors subject to 10 CFR Part 50 or 10 CFR Part 52. The regulations at 10 CFR Part 100 s pecify that the requirements are for stationary reactors, so the NRC staff wo uld need to evaluate regulatory applicability (e.g., could mobile reactors be consid ered stationary at each site of operation?) and how mobile micro-reactor applicants wou ld meet the underlying purpose of the 10 CFR Part 100 requirements without knowing specific deployment sites in advance.36

(2) NEPA requires Federal agencies to evaluate the impacts of p roposed Federal actions on the human environment. The regulations in 10 CFR Par t 51 form the basis for the NRC's NEPA compliance and direct the NRC staff on how t o perform environmental reviews. As a Federal agency, the NRC must assess the environmental effects of proposed actions before making decisio ns. In order to be able to authorize operation of a mobile micro-reactor on an as-needed, where-needed basis without preapproved sites, the environmental revie w would potentially need to cover deployment at any location within the United Stat es. The NRC staff would need to further evaluate the feasibility of performing su ch a review and possible appropriate ways to meet the NEPA requirements, perhap s through bounding site parameters or other means.

Section 4 of this enclosure discusses transportation requiremen ts for fueled micro-reactors. The transportation considerations for mobile micro-reactors would b e the same as those described in section 4 for factory-fabricated modules loaded with fuel th at had been operated for some time (such as operational testing at a factory or operation at a deployment site). Except for package approvals under 10 CFR 71.41(d), transportation package certifications are issued in accordance with the requirements in 10 CFR Part 71 and can be u sed to support an unlimited number of transportation events without additional licensing. H owever, if the Commission does not approve the NRC staffs proposal in Option 1b of this paper for the use of features to preclude criticality, the reactor would be considered to be in operation when loaded with fuel, including during transit. The current regulatory framework does not contemplate licensing reactors that are in operation while in movement, nor has the NRC developed criteria to assess the safety of reactors in operation while in movement.

Near Term Strategy

Besides the proposals for the use of features to preclude criti cality described in this paper, the NRC staff does not intend to act in the near-term to address ch anges to the regulatory framework that may be needed to support mobile micro-reactor li censing.

Next Steps

The NRC staff will monitor developments in the commercial secto r related to deployment models and the demand for commercial mobile micro-reactors. If developers place additional

36 For 10 CFR Part 52 design certifications, the NRC staff bases its safety review of the standard design on postulated site parameters that would bound a number of sites. However, an applicant for a particular site performs site investigations and analyses to establish that the characteristics of its site falls within the postulated site parameters, and the NRC staff reviews this information for each site-specific application.

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focus on commercial mobile micro-reactor deployment, the NRC st aff will assess the need for changes in the regulatory framework and Commission direction. D epending on the interest in applications for commercial mobile micro-reactors for particula r uses (such as disaster relief),

the NRC staff will consider the need to engage other Federal Go vernment agencies and the need to develop a new regulatory framework.