ML23236A597
| ML23236A597 | |
| Person / Time | |
|---|---|
| Issue date: | 08/29/2023 |
| From: | William Kennedy NRC/NRR/DANU/UARP |
| To: | |
| References | |
| Download: ML23236A597 (32) | |
Text
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Enclosure to NRC Staff Prepared White Paper Micro-Reactor Licensing and Deployment Considerations:
Fuel Loading and Operational Testing at a Factory August 2023 Draft - Released to Support ACRS Interaction
THIS NRC STAFF WHITE PAPER HAS BEEN PREPARED AND IS BEING RELEASED TO SUPPORT INTERACTIONS WITH THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS). THIS PAPER HAS NOT BEEN SUBJECT TO NRC MANAGEMENT AND LEGAL REVIEWS AND APPROVALS, AND ITS CONTENTS SHOULD NOT BE INTERPRETED AS OFFICIAL AGENCY POSITIONS.
Technical, Licensing, and Policy Considerations for Factory-Fab ricated Micro-Reactors
This enclosure includes various topics that are related to the licensing and deployment of factory-fabricated micro-reactors (and other technologies, as a ppropriate). Some of these are topics raised by developers through formal pre-application enga gement with the U.S. Nuclear Regulatory Commission (NRC, the Commission) 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 (ML20129J985), and in the context of small modular reactors or non-light-water reactors, but are revisited here with the attri butes of factory-fabricated micro-reactors and related deployment models in mind. The NRC staff will address design-specific issues on a case-by-case basis.
This enclosure also includes the NRC staffs near-term strategi es and 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 Comm ission direction. These are included to provide factory-fabricated micro-reactor developers and potential applicants with awareness of approaches that are available now. The next steps are focused on longer-term approaches that may involve additional Commission engagement in the future, as appropriate.
- 1. Considerations Related to Initial Fuel Load and Authorizatio n to Operate at the Deployment Site for Reactors that Arrive Pre-Loaded with Fuel
Deployment Model Considerations
Deployment strategies that include loading fuel at a factory wo uld result in fueled factory-fabricated modules arriving at the deployment site. However, several requirements in the Atomic Energy Act of 1954, as amended (AEA) and 10 CFR Parts 50 and 52 that are related to public notifications, the opportunity for hearing, and authorization t o operate the facility under a combined license are premised on the initial loading of fuel at the deployment site. For example, AEA Section 189a.(1)(B)(i) requires the following:
Not less than 180 days before the date scheduled for initial lo ading of fuel into a plant by a licensee that has been issued a combined constructio n permit and operating license under section 185b., the Commission shall pub lish in the Federal Register notice of intended operation. That notice shal l provide that any
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 2
person whose interest may be affected by operation of the plant, may within 60 days request the Commission to hold a hearing on whether the fa cility as constructed complies, or on completion will comply, with the ac ceptance criteria of the license.
In the case where a factory-fabricated module arrives at a depl oyment site for which a combined license has been issued, the fabricator would have loaded fuel at the factory under its license.
Therefore, at the deployment site there would not be initial loading of fuel into a plant by a licensee that has been issued a combined construction permit an d operating license. This entry condition to the requirements in AEA Section 189a.(1)(B)(i) wou ld not be satisfied as written, but it would be inconsistent with the laws purpose for the Commiss ion not to publish the notice of intended operation and opportunity for the public to request a hearing on conformance with the acceptance criteria in the combined license for the deployment site. Thus, the NRC should find the closest analogue to initial fuel load by the combined licen se holder that also fulfills the underlying purpose of the law. 1
In addition, AEA Section 185b. states, in part, that [f]ollowi ng issuance of the combined license, the Commission shall ensure that the prescribed inspections, te sts, and analyses are performed and, prior to operation of the facility, shall find that the pr escribed acceptance criteria are met.
The Commission has historically considered fuel loading to be p art of operation, meaning that a reactor that arrives at the deployment site loaded with fuel wo uld be in operation before the Commission makes the required finding that the prescribed accep tance criteria are met. As described in the options in this paper, the NRC staff proposes the use of features to preclude criticality to define fuel load under these circumstances as no t being in operation and considering whether the removal of the criticality preclusion f eatures would be the best analogue to the initial loading of fuel to which AEA Section 18 9a.(1)(B)(i) refers. If approved by the Commission, features to preclude criticality would avoid th e situation in which a fueled reactor arrives at the deployment site and is already considere d to be in operation.
The regulations in 10 CFR Part 52, Subpart C contain requiremen ts related to these provisions of the AEA. In relation to AEA Section 185b., the regulations i n 10 CFR 52.103(g) require, in part, that [t]he licensee shall not operate the facility until the Commission makes a finding that the acceptance criteria in the combined license are met In re lation to AEA Section 189a.(1)(B)(i), the regulations in 10 CFR 52.103(a) require tha t [t]he licensee shall notify the NRC of its scheduled date for initial loading of fuel no later than 270 days before the scheduled date and shall notify the NRC of updates to its schedule every 30 days thereafter. There are other regulations that reference initial fuel loading, discusse d below. An applicant for or holder of a combined license could reques t exemptions from these regulations if the provisions, as written, cannot be reasonably complied with and the requested e xemptions comply with the AEA. License conditions may be needed to address any regulatory gaps. If, as proposed by the NRC staff in this paper, a factory-fabricated module that inclu des features to preclude criticality is not in operation when loaded with fuel, then the requirement s in AEA Section 185b. and 10 CFR 52.103(g) could be met as long as the module were not place d into operation (e.g.,
through the removal of features to preclude criticality) until after the Commission makes a
1 Section 189a.(1)(B)(v) of the AEA also refers to fuel load: The Commission shall, to the maximum possible extent, render a decision on issues raised by the hearing request within 180 days of the publication of the notice provided by clause (i) or the anticipated date for initial loading of fuel into the reactor, whichever is later. The NRC staff understands anticipated date for initial loading of fuel into the reactor in context to refer to the scheduled date for initial loading of fuel by the COL holder that is discussed in AEA Section 189a.(1)(B)(i).
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 3
finding that the acceptance criteria in the combined license ar e met. This could also address reactors arriving at a deployment site for which a construction permit has been granted under 10 CFR Part 50 but not an operating license.
Near Term Strategy
If the Commission approves the NRC staffs proposed use of feat ures to preclude criticality, then a factory-fabricated module will not be in operation durin g transportation to the deployment site or upon arrival. If directed by the Commission under Optio n 1b in this paper, the NRC staff could use the removal of these features as an analogue to initi al fuel loading by the combined license holder for the purposes of the notification and opportu nity for hearing in AEA Section 189a.(1)(B)(i) and as the equivalent of the commencement of ope ration for the purpose of AEA Section 185b. The NRC staff notes that the removal of features to preclude criticality and initial loading of fuel are both distinct actions performed b y the combined license holder that put a fully constructed utilization facility in a position to s ustain a nuclear chain reaction. In both cases, the utilization facility is incapable of sustaining a nu clear chain reaction (for lack of sufficient reactivity) before operation is authorized. Operatio n is not authorized until the Commission finds that the prescribed acceptance criteria are me t, and the public is given an associated opportunity to request a hearing on whether the faci lity satisfies the acceptance criteria.
For fueled reactors being deployed on sites for which a constru ction permit has been granted under 10 CFR Part 50, the NRC staff could also use removal of f eatures to preclude criticality as an analogue to initial fuel loading and the beginning of operat ion and would authorize removal of the features only after an operating license had been granted.
There are numerous other regulations including 10 CFR 50.47, 10 CFR 50.54, 10 CFR 50.55a, 10 CFR 50.71, 10 CFR 50.75, 10 CFR 50.120, 10 CFR Part 50 Appen dix E, 10 CFR Part 50 Appendix J, 10 CFR 52.99, and other provisions in 10 CFR 52.103 where initial fuel load is used as a milestone. In these cases, the requirements to be implemen ted at fuel load are not linked to specific language in the AEA. Therefore, the NRC staff could use the removal of features to preclude criticality as an analogue to initial loading of fue l and could implement this approach through exemptions and license conditions, as appropriate, in t he near term.
Next Steps
Depending on Commission direction on Option 1b in this paper, t he NRC staff will consider whether additional actions are warranted related to initial fue l load and authorization to operate at the deployment site for reactors that arrive pre-loaded with fuel
- 2. 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. Factory-fabricated micro-reactors t hat are of a self-contained design with the nuclear and balance-of-plant systems in one or a few containers that are fully fabricated at the factory would likely require only simple cons truction activities at the deployment site (e.g., pouring a small concrete pad on which to place the container housing the reactor).
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 4
This type of reactor might be ready for operation within days t o weeks of receipt of the construction permit or combined license for the deployment site if construction begins immediately after license issuance. Factory-fabricated micro-re actor designs that have a core module design and would require more complex construction activ ities at the deployment site (but still much simpler than construction activities for large light-water reactors), such as erecting a reactor building and installing power conversion equ ipment, might be ready for operation within a few months of the start of construction. In either case, a key aspect of factory-fabricated micro-reactor deployment models is the ability to mo ve a factory-fabricated module from the factory to the deployment site and place it into opera tion as a nuclear power plant in a much shorter time than it takes to construct a large light-wate r reactor at the intended site of operation.
Factory-fabricated micro-reactors may be licensed at the deploy ment site under either 10 CFR Part 50 or 52. In both cases, a mandatory hearing would be held as part of the process for issuing the authorization to construct the reactor at the deplo yment site. Factory-fabricated micro-reactors would likely have standardized designs that may be described in a referenced 10 CFR Part 52 design certification or manufacturing license, w hich could reduce the scope of NRC review for deployment site li censing. Also, any final safety findings on final design information in a construction permit application would be incor porated in the permit in accordance with 10 CFR 50.35(b) and subject to the backfitting requirements in 10 CFR 50.109.
Under a 10 CFR Part 50 approach, an additional opportunity for hearing is required by AEA Section 189a.(1)(A) in conjunction with issuance of the facilit y operating license. The regulations in 10 CFR 2.309(b)(3) provide a 60-day opportunity to request a hearing (although AEA Section 189a.(1)(A) requires 30 days notice). The potential scope of s uch a hearing would be the entirety of the operating license application, but the hearing scope would be reduced to the extent the operating license application references an earlier NRC license or approval providing finality on the matters resolved therein, such as a manufacturi ng license. Subparts C and L and Appendix B of 10 CFR Part 2 provide the rules of general applic ability, procedures, and model milestones for such a hearing. If no hearing is requested, the Commission could issue the operating license immediately upon the closure of the 60-day he aring request period, provided all other requirements are met. In the case that a hearing is r equested, the issuance of the license would have to wait until a Commission decision to not g rant the hearing request or else the completion of the hearing if the request is granted, 2 either of which could take many months according to the procedures and model milestones in 10 CFR 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 nuc lear power reactor with a supporting environmental impact statement (EIS), the regulation s in 10 CFR 51.20(b)(2) and 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 for prep aration and publication of 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 (NEPA) was amended in June 2023, to in clude a new requirement to issue EISs within 24 months. Recent improvements in the environ mental review process along
2 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 without a hearing, but upon thirty days notice and publication once in the Federal Register of its intent to do so (emphasis added).
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 5
with standardized, relatively simple reactor designs, and small reactor sites could reduce the time to complete a supplemental EIS to less than 24 months, wit h further process improvements planned. This is in contrast to the environmental review for a combined license, in which the environmental review is completed at the time of issuance and a supplement to the EIS is not required in connection with the 10 CFR 52.103(g) finding and th e authorization to operate and would not contribute to the deployment timeframe.
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 (or removal of features to pre clude criticality if the Commission approves Option 1b in this paper) by the combined license holde r and specify a 60-day period for the opportunity to request a hearing. The potential scope o f such a hearing would be limited to the inspections, tests, analyses, and acceptance criteria (I TAAC) included in the combined license.
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, the licensee shall notify the NRC of its scheduled date for initial loading of fuel no later than 270 days before the scheduled dat e.
- The regulations in 10 CFR 52.99 include requirements on the ti ming of licensee notifications of ITAAC closure and completion and NRC publicati on of related notices. In particular, 10 CFR 52.99(c)(3) provides that if the licensee ha s 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 fuel, then the licensee m ust provide an uncompleted ITAAC notification no later than 225 days before the scheduled initial fuel load date to describe how the licensee will complete the uncompleted ITAAC. The intent of this requirement, in part, is to ensure that information related to ITAAC closure is available to the NRC staff and the public at the time the Commission publish es 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 require ments 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, render a decision on issues raised by the hearing request within 180 days of the pub lication 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 that can be deployed in a matter of days to a few months, these timeframes will likely result in delays in entering the reactor into operation. If the licensee were to no tify the Commission of the intended date of initial fuel load upon receipt of the combined license, that notification would start the 270-day period. By the end of that 270-day period, the Commissi on would normally complete the hearing, if requested and granted, and be able to determine whether it can make the 10 CFR 52.103(g) finding. Based on the 270-day notification by the licensee, the Commission
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 6
would have 90 days to publish the notice of intended operation and 60-day opportunity to request a hearing required by AEA Section 189a.(1)(B)(i). Publi cation of this notice would start a 180-day period to normally complete the hearing process per AEA Section 189a.(1)(B)(v).
However, as discussed in the NRCs ITAAC hearing procedures, th e Commission has established a goal to publish the notice of intended operation 210 days before scheduled fuel load.
Notwithstanding the Commissions obligations under AEA Section 189a.(1)(B)(i), the Commission has previously established that a licensee may under certain conditions begin operation before the scheduled date of initial fuel loading sub mitted to the Commission under 10 CFR 52.103(a). The NRC stated at 81 FR 43273 in the publicat ion of the final ITAAC hearing procedures that the licensee can, consistent with 10 CFR 52.10 3(a), move up its scheduled fuel load date after the notice of intended operation is publis hed. Such a contraction in the licensees fuel load schedule would have no effect on the heari ng schedule, but as a practical matter, the NRC would consider such a contraction in the licens ees schedule as part of its process for making the 10 CFR 52.103(g) finding and the adequat e protection determination for interim operation. In its Comment Summary Report - Procedures for Conducting Hearings on Whether Acceptance Criteria in Combined Licenses Are Met (ML16 167A464), Section 5, Hearing Tracks and Schedules, Subsection G, Contraction of F uel Load Schedule, the NRC stated that in the absence of a hearing or if the hearing issue s are resolved early in favor of the licensee, the licensee will be allowed to operate if and after the 10 CFR 52.103(g) finding is made. The NRC also stated in the Comment Summary Report that if a hearing is held and has not been completed, but the NRC staff has made the 10 CFR 52.10 3(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 load3 and the 10 CFR 52.99(c) ITAAC closure notifications and uncomp leted ITAAC notifications for all ITAAC immediately upon receipt of the combined license. If the combined license applicant intended to do this, it should inform the NRC staff of its intention in the combined license application or by other means so that the NRC staff cou ld make necessary arrangements to prepare the notice of opportunity for hearing. Consistent with the NRCs experience with the Vogtle ITAAC proceedings, the NRC could mak e publicly available the
3 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.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 7
uncompleted ITAAC notification and publish the notice of intend ed operation within about 15 days after receipt of the uncompleted ITAAC notification.
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 in ideal circumstances, which m ay be driven primarily by the statutory requirement in AEA Section 189a.(1)(B)(i) that the no tice of intended operation provide that any person whose interest may be affected by oper ation of the plant, may within 60 days request the Commission to hold a hearing. In cases wh ere a hearing is requested, the minimum timeframe would be extended in accordance with the final ITAAC hearing procedures, i.e., the 10 CFR 52.103(g) finding might be issued (1) after the Commissions decision on the hearing request if the request is denied, (2) a fter a decision allowing interim operation if the hearing request is granted and the requirement s for interim operation are met, or otherwise, (3) after the presiding officer has issued the decis ion after hearing.
If a hearing is requested, the regulations and ITAAC hearing pr ocedures allow the licensee and the NRC staff 25 days to answer the hearing request and establi sh a milestone of 30 days after the answers for a Commission ruling on the hearing request. If the Commission does not grant a hearing request, this would add 55 days to the minimum deployme nt timeframe 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 would then complete its final safety evaluation durin g the 60-day period of the opportunity to request a hearing. This assumes that the final d esign, site-specific issues, technical specifications, and operational programs would have b een reviewed and approved during the construction permit proceedings or other prior appro vals (e.g., topical reports) and the operating license application doesnt introduce any deviati ons. In cases where a hearing was not requested and all other requirements are met, the deplo yment timeframe could be shortened to approximately 95 days, which accounts for 30 days to perform an acceptance review and docket the application, 5 days to publish the notice of opportunity for hearing, and 60 days for the opportunity to request a hearing. In cases where a hearing is requested, the timeframe will be extended in accordance with the regulations i n 10 CFR Part 2. However, under the 10 CFR Part 50 process for issuing a facility operati ng license, the record of decision cannot be issued until both the safety and environmental review s are completed. Characteristics of factory-fabricated micro-reactors, such as standardized desi gns and relatively small site footprints with limited construction activities at the deployme nt site, may allow for the NRC staff to complete the required supplement to the EIS in less than 24 months under the current process. In accordance with SRM-SECY-21-0001, Rulemaking Plan Transforming the NRCs Environmental Review Process (ML22109A171), after completing s everal 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
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 8
subcategories of license applications presently falling within the scope of 10 CFR 51.20(b), and present options to the Commission. If such a proposal was devel oped and approved, environmental assessments could be used both for the constructi on permit and the operating license at the deployment site, which would require appropriate changes to 10 CFR Part 51 or exemptions. The use of environmental assessments instead of EIS s could substantially shorten the timeline such that the deployment timeframe could be dictat ed by the operating license contested hearing process.
Despite the differences in the requirements for opportunities f or hearings for issuance of an operating license or making the 10 CFR 52.103(g) finding, the N RC staff hasnt found any reasons that either the 10 CFR Part 50 or Part 52 processes wou ld necessarily result in significantly different deployment timeframes, except for the t ime needed to issue an EIS supplement to support issuance of an operating license. Ultimat ely, it will be up to an applicant to consider the differences in the two licensing processes and decide which one is better suited to its 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 Commission direction on the options presented in this paper for features to preclude criticality, fuel loading, and o perational testing; the characteristics of factory-fabricated micro-reactors; and further stakeholder i nput. If warranted, the NRC staff will propose an update to the final ITAAC hearing procedures fo r Commission consideration and consider rulemaking options, as appropriate, to ensure that hea rings would not result in unnecessary delays to operation of factory-fabricated micro-rea ctors. The NRC staff is considering whether there are opportunities to streamline the p rocess for preparing and publishing the supplement to the EIS if licensing under 10 CFR Part 50 is pursued by a potential applicant.
- 3. 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. 4 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.
4 The NRC staff is aware that the designs of factory-fabricated micro-reactors 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. In this case the, the fabricator would have amended its manufacturing license or obtained a new manufacturing license for the new design. Likewise, the deployment site licensee would have to amend its permits and licenses to account for the new design or obtain new permits and licenses.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 9
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 (SMRs) considered in SECY-11-0079 is that the replacement factory-fabricated micro-reactor modules are not in tended to be operated at the same time as the module it is intended to replace. Each module would operate for limited time (generally less than 10 years) and then would be removed from s ervice and replaced. Upon removal from service, the deployment site licensee would instal l features to preclude criticality to take the module out of operation and store the module on site p rior to decommissioning or shipment to a decommissioning facility or a refurbishment and r efueling facility.
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 for all factory-fabricated modules, including replacement modules, anticipated to be operated at the deployment site. The application would also specify the number of modules that could be present at the site simulta neously and operated simultaneously. The combined license application would also inc lude any permanently installed site-specific features such as power conversion systems or stru ctures, if applicable. Under this approach, each module would receive its own combined license; h owever the licenses would be issued concurrently. This would be similar to Alternative 3: Individual 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 life of t he 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 associa ted opportunity for hearing) before the first module and every subsequent replacement module would be authorized to begin operation under its combined license. There are potential timin g impacts associated with completing ITAAC and potential ITAAC hearings for each replacem ent module, but the license application, safety review, mandatory hearing, and opportunity for contested hearing on the combined license issuance would only be conducted once for all combined licenses for the modules anticipated to be operated at the facility. The NRC sta ff expects that licensees would
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 10
know in advance of the need for the replacement module and ther efore could plan accordingly to minimize potential impacts on plant downtime caused by the t ime required for ITAAC completion, a potential ITAAC hearing, and the Commissions fin ding 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 onsite, 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. D epending on the deployment facility and the licensees plans for operation, the final safe ty analysis report may need to 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 there were no other operating modules at the site.
Under 10 CFR Part 50, the NRC could issue one construction perm it covering the construction of the facility (i.e., the factory-fabricated module and the on -site balance-of-plant) including all replacement modules anticipated to be deployed at the site. Iss uance of the construction permit would involve a mandatory hearing, opportunity for contested he aring on the construction permit issuance, and an EIS that would consider all of the modules ant icipated to be deployed at the site. The NRC could then issue separate operating licenses to a uthorize operation of each replacement module after they were installed at the deployment site. To the extent that the first module and all replacements would have the same standard design and any permanent on-site structures or features would have been approved with the issuan ce of the operating license for the first module at the deployment site, the NRC staff could le verage the safety review that had already been completed for subsequent operating licenses. Issua nce of the operating license would require the NRC to provide an opportunity for the public to request a hearing and to publish a supplement to the final EIS for the construction perm it as required by 10 CFR 51.20(b)(2) and 51.95(b). Per 10 CFR 51.95(b), the supplement t o the EIS would only cover matters that differ from the final EIS for the construction per mit or that reflect significant new information concerning matters discussed in that final EIS.
Under the 10 CFR Part 50 approach, the NRC staff expects that d epending on the design of the factory-fabricated module, the potential downtime could be mini mized by the operating license application being submitted well in advance of removing the cur rently operating module from service, which should allow the EIS to be supplemented and the contested hearing process to be completed. However, the process for supplementing the EIS an d the opportunity for a contested hearing to support issuing the operating license for the replacement module introduce the potential for new issues to arise. This is in contrast to t he approach using 10 CFR Part 52 under which the design and environmental reviews are completed and given finality upon issuance of the combined licenses.
Near Term Strategy
The NRC staff intends to use the existing regulations in 10 CFR Parts 50 and 52, as informed by SECY-11-0079 and the approach described above, for licensing replacement of factory-fabricated modules. Factory-fabricated micro-reactor developers and potential applicants will need to consider factors such as the number of modules that wil l be onsite at one time, the expected operational states of the onsite modules, replacement frequency, and others when deciding whether to apply for licensing of replacement modules under 10 CFR Part 50 or 52.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 11
Next Steps
The NRC staff intends to continue to engage with stakeholders i n pre-application activities and public meetings to further understand and assess planned deploy ment 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 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.
- 4. Autonomous Operation and Remote Operation
Deployment Model Considerations
During recent pre-application interactions with the NRC staff, significant interest has been expressed by micro-reactor developers regarding the inclusion o f autonomous5 and remote operational characteristics within their proposed designs. A re mote operational model tends to center around the minimization of the numbers of both operators and other categories of facility staffing at the facility site, while an autonomous operational model would seek to eliminate reliance upon the use of operators. For the purposes of this pa per, autonomous systems are considered those able to perform their task and achieve their functions independently (of the human operator), perform well under significant uncertainties f or extended periods of time with limited or nonexistent communication, with the ability to compe nsate for failures, all without external intervention."6 In the case of remote operations, the objective is to relocate staff to a centralized location and operate some reactors remotely. Such a pproaches differ dramatically from the current paradigm of commercial nuclear plant operation s in which operators are required by 10 CFR 50.54(k) and (m) to maintain a continual ons ite presence in control rooms.
Furthermore, such operational approaches may entail the elimina tion of a main control room at the facility, thereby requiring the evaluation of requested exe mptions from relevant regulations, such as 10 CFR 50.34(f)(2)(iii). 7 Thus, proposed deployment models that would involve relocating operators offsite or eliminating them entirely prese nt a number of significant differences from traditional reactor designs and licensing para digms.
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
5 As used within the context of this paper, the terms automation and autonomous have distinct meanings. Automation refers to automated processes. The term automated, in turn, is defined herein as the independent performance of tasks via the application of technology and absent continuous input from an operator. It should be noted that the proposed 10 CFR Part 53 rulemaking would define automation as a device or system that accomplishes (partially or fully) a function or task.
6 M. R. Endsley, From here to autonomy: lessons learned from human-automation research, Human factors, vol. 59, no. 1, pp. 5-27, 2017.
7 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.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 12
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 be able to take ad vantage of being co-located with the reactor facility in order to receive sensory feedback (e.g., noise, vibrations, local observation of conditions, etc.) that would serve to augment the informatio n otherwise provided to them through the plants instrumentation and control interfaces. Suc h information can be very useful in conditions of instrumentation and control failures, particul arly in instances where highly automated systems are involved. Autonomous and remote operation s approaches represent a shift in operator capabilities that need to be carefully consid ered and are areas in which the NRC staff has been working to develop the needed HFE tools as p art of broader efforts at developing an HFE framework for the review of advanced reactor designs.8, 9
Autonomous 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 to be the ability to operate w ith complete independence from human control, while automation refers to the machine exec ution of what were formerly human tasks.10 Thus, autonomous operation may not rely on high levels of auto mation and could be achieved via simplicity (e.g., reliance upon inherent safety characteristics and robust passive systems) of an advanced reactor design. 11 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.12 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. 13 Such functionality (i.e., the incorporation of artificial inte lligence)
8 See, e.g., SECY-20-0093.
9 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 Parasuraman, R., & Riley, V. (1997). Humans and automation: Use, misuse, disuse, abuse. Human Factors, 39(2), 230-253.
10 NUREG-2261, Artificial Intelligence Strategic Plan Fiscal Years 2023-2027 (ML23132A305),
includes a notional framework for artificial intelligence and autonomy levels in commercial nuclear activities. However, it should be noted that autonomy is generally considered within the context of artificial intelligence (e.g., machine learning) by that document. For that reason, this paper relies upon the definition of autonomous that was presented earlier during the discussion of autonomous systems in section four of this enclosure.
11 The concept of autonomy (particularly where safety performance is concerned) potentially not being achieved by complex automation but rather via 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.
12 Gaining clarity on industry deployment concepts for autonomous operation will be an area of focus under the Next Steps portion of this section.
13 National Academies of Sciences, Engineering, and Medicine. 2022. Human-AI Teaming: State-of-the-Art and Research Needs. Washington, DC: The National Academies Press.
https://doi.org/10.17226/26355.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 13
will present new considerations for both developers and the NRC staff because it will introduce the potential for automation to potentially take actions other than those that were originally assumed during the design and licensing processes. Furthermore, the acceptability of potentially allowing AI-driven automation to control safety-sig nificant operations (e.g., those needed to mitigate accidents) represents a presently unresolved matter and an area of ongoing discussion amongst designers, researchers, 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 where the micro-reactor is the primary sour ce of power, as well as those where it supplies the grid in parallel with other generation as sets. Such uses would benefit substantially from the ability to let grid demand directly chan ge reactor power without a human operator serving as an intermediary, such as by permitting grid control centers staffed by non-NRC-licensed individuals to directl y control the electrical output of the micro-reactor facility.
However, load-following operation in which a non-licensed indiv idual modifies the power level of a nuclear reactor is precluded by the current NRC regulations. As background, AEA Section 11r. defines operators as individuals who manipulate the contro ls of utilization or production facilities. The AEA then mandates under Section 107 that indivi duals who operate controls must be licensed by the NRC. Notably, the AEA does not define w hat those controls consist of, thus affording the NRC the discretion to establish that def inition via regulation. Both 10 CFR 50.2 and 55.4 define these controls (when used within the con text of nuclear reactors) as consisting of apparatuses and mechanisms that directly affect t he reactivity or power level of the reactor when manipulated. From the inception of operator licens ing in 1956, manipulation of the controls of a utilization facility has been restricted to licen sed operators under 10 CFR 50.54(i).
The NRC staff anticipates micro-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 technological requirements wou ld be necessary to provide for the reliable and secure monitoring and control of one or more m icro-reactor units from a remote location. For example, commercial large light-water reactor lic ensees have historically credited human actions for performing certain time-critical operations t o meet accident analysis assumptions. Any suitable approach to remote operations would, logically, need to either provide a very high degree of assurance in the ability of opera tors to remotely accomplish such actions or, alternatively, eliminate reliance upon such actions for the achievement of safety functions. Two further considerations need to be taken into acc ount here as well. First, under a remote operations approach, the absence of operators on site ma y potentially remove any opportunity for local, backup actions should remote operations be unsuccessful for any reason.
Secondly, the viability of any remote operations approach would be predicated on the ability of developers to adequately address the needed cybersecurity consi derations for remote operations. This latter point will be explored in greater depth within this section.
The remote operation of commercial nuclear power plants, has no t yet been explored extensively from a practical implementation standpoint, and the refore there is a paucity of operating experience to draw from to inform future approaches t o remote operations. The feasibility of an unattended light-water reactor design was stu died in the early 1960s and a determination was reached that the concept was dependent on whe ther safety systems could be designed to a level of reliability that was high enough to p reclude the need for regular
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 14
maintenance, thereby necessitating a relatively simple design. 14 At present, areas of regulatory guidance needed to further develop the concept remain. Centrall y, these include resolving issues regarding what design attributes would be necessary to s upport a safety determination for a remotely operated reactor facility. For example, the pote ntial for loss of remote control capability may warrant requiring remotely operated reactors to meet technological criteria comparable to those proposed for self-reliant-mitigation facil ities under 10 CFR Part 53 based upon a need to robustly demonstrate safety in the absence of an y opportunity for human intervention.
Instrumentation and Control
A systematic, comprehensive assessment of potential internal an d external hazards, including the potential for human-induced events, and their consequences must be performed as part of the plants safety analysis. Such an assessment should factor i n any potential hazards stemming from the introduction of autonomous and/or remote oper ational instrumentation and control (I&C), which must be identified, analyzed, and appropri ately addressed (e.g., prevented or mitigated) as part of the I&C design for safety. When I&C sy stems are relied upon to satisfy the overall nuclear power plant performance objectives, the des ign of the risk significant I&C systems must meet applicable regulations for safety. The I&C de sign, used for functions such as sensing, control, display, and monitoring of the plant, shou ld be sufficiently reliable and robust commensurate with its safety significance as required by the regulations.
The applicable regulations, such as those under 10 CFR Part 50 or 52, for I&C are generally technology inclusive and perform ance-based as they primarily re quire the adequate demonstration of reliability (e.g., testing, surveillance, fail -safe design, and quality) and robustness (e.g., redundancy, independence, diversity, defense-in-depth, and deterministic behavior, and qualification) independent of the I&C technologie s 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 Design Review Guide (DRG) for I&C of Non-Li ght-Water Reactors (ML21011A140), which provides guidance for the NRC staff to use in reviewing the I&C portions of applications for advanced non-light-water reactors that foll ow Regulatory Guide (RG) 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Perfor mance-Based Methodology to Inform the Licensing Basis and Content of Applications for Lice nses, Certifications, and Approvals for Non-Light-Water Reactors (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 limiting maintenance access. Furthermore, I&C equipment 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. Corrosivity as well as high-temperature coolants will affect the operative lifespan of I&C equipment as well as contribute to measurement uncertainty. Micro-reactors that operate with therm al or fast neutron spectra may also have different sensor/actuator requirements due to a wide variety of fuels, higher-operating
14 Rosenthal, M.W., et al., The Feasibility of an Unattended Nuc lear Power Plant, Oak Ridge National Laboratory Report, ORNL-2985, August 1960.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 15
temperatures, and flexible operation modes. These issues may po se challenges for I&C equipment that may need to be addressed in additional regulator y guidance.
Cybersecurity
The current power reactor cybersecurity requirements in 10 CFR 73.54 extend to 1) digital computer and communication systems and networks that are associ ated with safety-related, important-to-safety, security, and emergency preparedness (SSEP ) functions and 2) support systems and equipment that, if com promised, could adversely imp act SSEP functions. To implement the requirements, licensees identify critical digital assets (CDAs) that must be protected against cyberattacks.10 CFR 73.54 does not specifical ly address autonomous or remote operations; however, the performance-based nature of the regulation supports both non-autonomous, autonomous, and remote operations through applicati on of appropriate cybersecurity considerations in a licensees 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 CFR Part 52 would b e required to protect the security posture of an autonomous and/or remotely-operated syst em with the same level of assurance applied to non-autonomous and locally-operated digita l computer and communication systems and networks.
Data communication pathways would be within the scope of digita l computer and communication systems and networks required to be protected und er 10 CFR 73.54(a) and would require protection against cyberattacks, up to and includ ing the design basis threat as described in 10 CFR 73.1. Depending on the level of autonomy, m icro-reactor designers would likely propose employing data connections using wired, wireless, or a combination of both pathways to communicate with critical systems and CDAs. Technic al security controls in current nuclear power plant licensee cybersecurity plans prohibit the u se of wireless technology for CDAs associated with safety-related and important-to-safety fun ctions. A defensive computer security architecture that employs any type of remote access co uld be vulnerable to cyberattacks, such as unauthorized remote access, complete deni al of service, and/or denial of authorized remote access functions. The level of autonomy, remo te operations, and remote monitoring are important aspects in understanding the associate d cybersecurity risks. A defensive computer security architecture that incorporates remo te operation technology would have to be more complex than those of current licensees to ensu re the confidentiality, integrity, and availability of digital computer and communication systems and networks associated with SSEP functions.
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-I SG-2023-03, Development of Scalable Human Factors Engineering Review Plans (ML22266A072, nonpublic). The draft guidance describes a method for scaling the scope and depth 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 NRC staff HFE review efforts considering fac tors such as risk insights and the unique characteristics of the design or facility operation (e.g., remote 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 would intend to use the general methods described in this guidance to scale HFE reviews of micro-reactor application s submitted under 10 CFR Part 50 or 52.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 16
For near-term license applications that include proposals for r emote or autonomous operation, the NRC staff would use available guidance to assess compliance with the applicable 10 CFR Part 50 or 52 regulations and applicant requests for exemptions, as needed.
Remote Operations
Appendix A to 10 CFR Part 50 contains the general design criter ia (GDC), which establish the minimum requirements for the principal design criteria (PDC) fo r water-cooled nuclear power plants similar in design and location to plants for which cons truction permits have been issued by the Commission. Appendix A also establishes that the GDC ar e considered to be generally applicable to other types of nuclear power units and are intend ed to provide guidance in determining the PDC for such other units. GDC 19, Control room, requires (for water-cooled reactors), in 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 ma intain it in a safe condition under accident conditions, including loss-of-coolant accidents. Regulatory Guide (RG) 1.232, Revision 0, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors, describes the NRCs guidance on how the GDC in Appendix A, Gen eral Design Criteria for Nuclear Power Plants, of 10 CFR Part 50 may be adapted for non -light-water reactor designs.
For nuclear reactors for which the GDC are requirements, the NR C staff would need to determine whether proposed PDCs for remote operation meet GDC 1 9. As noted in RG 1.232, the GDC in 10 CFR Part 50, Appendix A are not regulatory requir ements for non-light-water reactor designs but provide guidance in establishing the PDC fo r non-light-water reactor designs.
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 would intend to a pply this requirement to any facility from which a nuclear reactor can be operated remotely. In addition, the NRC staff would need to determine whether providing a facility from which the r eactor can be operated remotely would preclude the need for an on-site control capability beyon d that which would otherwise be provided under the 10 CFR Part 50, Appendix A, GDC 19 requireme nt for equipment at appropriate locations outside the control room with (1) a desig n capability for prompt hot shutdown of the reactor, including necessary instrumentation an d controls to maintain the unit in a safe condition during hot shutdown, and (2) a potential capab ility for 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 on-site licensed operator staffing that meets the c urrent staffing requirements.15 In 2005, the NRC staff published NUREG-1791, Guidance for Assessi ng Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Require ments Specified in 10 CFR
15 Staffing operators both on site and at a remote operations faci lity would not be an economical long-term strategy but is an option that licensees m ight exercise if it is deemed a viable pathway toward developing a performance-based case for r equesting an exemption from on-site staffing requirements at some point following init ial plant start-up.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 17
50.54(m).16 The NRC staff developed the guidance to support reviews of sta ffing for advanced reactors in which the concept of operations differed from that of the large light-water reactors upon which the current staffing regulations are based. The NRC staff would intend to evaluate staffing exemptions requests associated with applications propo sing remote operations using NUREG-1791 as the guidance addresses not only potential reducti ons in staffing but also the possibility of operation from remote facilities and portable de vices.
The NRCs requirements in 10 CFR Part 26 are applicable to, amo ng others, licensed operators at light-water reactor nuclear power plants currently licensed under 10 CFR Part 50 or 52.
However, 10 CFR 26.4(a), limits applicability to persons who a re granted unescorted access to nuclear power reactor protected areas and who, in the case of operators, perform the duties of operating or onsite directing of the operation of systems and components that a risk-informed evaluation process has shown to be significant to public health and safety. This leads to potential gaps in the coverage of these regulatory requirements. For example, a remotely located supervisory operator could potentially direct the opera tions of local personnel from a remote location; under such an arrangement, the supervisory ope rator would be neither operating nor onsite directing, and thus could bypass fitne ss-for-duty requirements. In developing these requirements, the NRC did not consider the pos sibility of remote operations where the reactor may be controlled by individuals who are not at the reactor site and therefore may not require unescorted access to a nuclear power reactors protected areas. Therefore, the NRC staff would need to address the limitations of the current 10 CFR Part 26 applicability language for license applications proposing remote operation of a nuclear reactor facility. The NRC staff would intend to address the limitations in a manner c onsistent with the proposed amendments to 10 CFR Part 26 in the 10 CFR Part 53 draft propos ed rule and the associated draft guidance in DG-5078, Fatigue Management for Nuclear Powe r Plant Personnel at Commercial Nuclear Plants Licensed Under 10 CFR Part 53 (ML222 64A109, nonpublic).
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, require that only licensed operators may manipulate controls, and require 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 52 for autonomous operations of nuclear reactors that, for e xample, would not include licensed operator staffing in a control room or would permit a non-licensed grid operator to change plant output (i.e., load following) would need to includ e requests for exemptions from these requirements, where relevant. Beyond this, additional reg ulatory implications will also need to be addressed, such as requirements for a control room a nd a remote shutdown capability. The NRC does not currently have guidance specific t o the evaluation of such requests. Additionally, as noted pr eviously, the AEA places certain mandates upon the NRC to license those individuals who will operate the controls of util ization facilities. However, as
16 More recently the NRC staff developed draft interim staff guid ance to augment NUREG-1791 to support NRC staff review of advanced reactor staffing p lans under 10 CFR Part 53 (i.e., DRO-ISG-2023-02, ISG Augmenting NUREG-1791, Guidance f or Assessing Exemption Requests from the Nuclear Power Plant Licensed Operat or Staffing Requirements Specified in 10 CFR 50.54(m), for Licensing Plant s under Part 53 (ML22266A068, nonpublic)).
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 18
described in the following section, Next Steps, the NRC staff has initiated research to more fully understand the extent to which reactor vendors intend to include autonomous operations in proposed concepts of operations for future applications and whe re there may be matters of importance or urgency for development of additional licensing g uidance.
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. Without a clear un derstanding of the architecture and how cybersecurity controls and safety protocols would be im plemented to ensure the security of a data communication pathway to the micro-reactor, it is difficult to determine risk, establish confidence, and credit important human actions needed to mitigate an event from compromise. Additionally, it is likely that protection against disruption or malicious control for micro-reactors will rely heavily on properly implemented DCSA i ncluding technical security controls - such as cryptography - rather than extensive use of physical security as is currently used in operating nuclear power plants. Furthermore, the level of autonomy and the capabilities of the autonomous technologies that would be used to replace hu mans for remote operations would be equally important to understand.
The regulations at 10 CFR 73.55 require, in part, that cybersec urity requirements must be implemented before fuel is allowed onsite (protected area). T he portable design of a factory-fabricated micro-reactor introduces a challenge to current cybe rsecurity regulations and guidance written for 10 CFR Part 50 and 52 plants. The NRC staf f would need to understand the security design elements that would be used to provide assu rance that systems and networks are protected from cyberattacks during fabrication, tr ansit, testing, and startup phases of the micro-reactors product development lifecycle.
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 review needed to address the deployment models, and develop the technical bases for any new guidance th at may be needed.
Regarding cybersecurity, the NRC staff has proposed as part of the 10 CFR Part 53 rulemaking, a new risk-informed, performance-based, technology-neutral cybe rsecurity requirements that apply a graded approach to advanced reactors for protection of digital computers, communication systems, and networks within the scope of the new requirement.
- 5. 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. As pr eviously discussed in this paper,
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 19
the use of features to preclude criticality would be needed in order to consider that the factory-fabricated module would not be in operation when loaded with fu el.
Shipment of a factory-fabricated module containing fuel would b e subject to the transportation requirements in 10 CFR Part 71. Shipment would also be subject to the requirements applicable to the licenses held by the fabricator for possession of the mo dules, special nuclear material, and any other radioactive material contained in the modules. As stated in this paper, the NRC staff assumes that the fabricator will obtain a manufacturing l icense under 10 CFR Part 52 for the factory-fabricated modules and that the manufacturing licen se would authorize possession of the modules at the factory. A license issued pursuant to 10 CFR Part 70 would also be required for possession of the special nuclear material in the fuel loaded in the module and a byproduct material license issued pursuant to 10 CFR Part 30 mi ght also be required if the module had been operated for testing and contained fission prod ucts.
Under a 10 CFR Part 52 manufacturing license, the licensee woul d be subject to the regulations in 10 CFR 52.153, 52.157, 52.167, in addition to 10 CFR Part 71 requirements. These requirements provide that the reactor is transported to a deplo yment site for which the licensee accepting the shipment has the proper licenses (e.g., a constru ction permit under 10 CFR Part 50 or a combined license under 10 CFR Part 52) and provide reas onable assurance that the reactor can be installed and operated at the deployment site. I n addition, the regulations in 10 CFR 52.157(f)(26)(iv) require that the application for a manufa cturing license include a description of the proposed procedures for shipment of the reac tor. These procedures, which may be different from the operating procedures submitted for pa ckage approval under 10 CFR Part 71, govern preparation of the reactor for shipping, perfor ming the shipment, and verification of the reactors condition upon receipt at the site to minimize the potential that the reactor arrives at its destination and is unable to operate within the parameters of the license at the deployment site.
The requirements in 10 CFR 70.20a and 10 CFR 70.20b contain pro visions for general licensees to possess special nuclear material and irradiated fu el during transport and storage incident to transport. In addition, 10 CFR Parts 70, 50, and 52 contain requirements for physical protection of fissile material in transport. The general licens e requirements in 10 CFR 70.20a and 70.20b also contain physical protection requirements to fol low the appropriate sections in 10 CFR Part 73, for material transport and storage incident to transport. However, the regulations for physical protection of special nuclear material in transport do not explicitly include requirements for special nuclear material that is loade d in a utilization facility 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.
Packaging and Transportation
The regulations in 10 CFR Part 71 contain requirements for Type B, 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 o f 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 history of t he uranium in the loaded fuel
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 20
and whether the module was operated at the factory for testing, 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 requirements in 10 CFR and t o the regulations of other agencies. A factory-fabricated module, whether shipped prior to operation or after operation, would be subject to the requirements in 10 CFR Parts 20 and 21, regardless of the contents or licensing pathways chosen by the Commission in response to this paper. In addition, the regulations in 10 CFR Parts 30, 50, 52, and 70 could apply, dep ending on the contents and parts in 10 CFR under which the reactor is licensed.
The NRC also co-regulates transportation with the Department of Transportation (DOT). The DOT regulates shipment of all classes of hazardous material, in cluding radioactive materials.
The shipper, and carrier are subject to the DOT regulations and depending on the components of the package, may be subject to more than one hazard class un der DOT regulations. DOT regulations authorize shipment of fissile material and radioact ive material in some NRC-approved packages under 49 CFR 173.415 and 173.416, respectivel y. However, if the NRC package approval options of either 71,41(c) or an exemption are utilized, the shipper must obtain a special permit issued by the DOT, as these package app rovals 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 for testing at a f actory and contains commercial grade uranium enriched to less than 20 weight percent in the ur anium-235 isotope would be classified as a Type AF package. However, if the package contai ns low-enriched fuel that includes reprocessed or downblended uranium, then the package w ould likely be classified as a Type BF package, depending on the quantity of impurities in the fuel and the A1 or A2 value17 for the mixture. Transport of a module that has been operated for t esting at a factory could call for a Type AF or Type BF package depending on the quantity of radionu clides generated during testing and the radionuclides initially present, e.g., if the f uel was reprocessed or downblended.
Based on the expected power level and duration of testing and t he time between the completion of testing and shipment, the applicant should determine the qua ntity of radionuclides that would be present at the intended time of shipment, using Appendix A t o 10 CFR Part 71, and determine whether the contents constitute a Type A or Type B qu antity of material. The regulations in 10 CFR 71.31 require the applicant for a package approval to provide a description of the contents and determine, based on the propose d contents, 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 and 71.59 and show that the dose rates for normal conditions of transport remain below the criteria in 10 CFR 71. 47. However, there would be no
17 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.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 21
containment or dose rate criteria after hypothetical accident c onditions for the Type AF package.
For Type BF packages, the package designer would be required to show that the package meets the hypothetical accident condition dose rate and contain ment criteria in 10 CFR 71.51 in addition to maintaining criticality safety and meeting 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 power level and duration of operation. The NRC staff estimates that after half a day of ful l power operation or equivalent, the module would likely contain greater than a Type A quantity of r adionuclides. Presuming the factory-fabricated module is operated at full power for more th an a single day, the applicable regulatory requirements would very likely be those for a Type B F package.
Near Term Strategy
The NRC and the DOT co-regulate transportation of radioactive m aterial, with the NRC regulating transportation for its licensees and issuing certifi cates of compliance for both fissile (Type AF and Type BF) and nonfissile (Type B) packages. 18 The DOT regulations in 49 CFR 173.416 and 173.417 authorize, among other things, shipment of any Type B or fissile material package approved by the NRC. The NRC staff intends to use the e xisting regulatory framework (primarily 10 CFR Part 71) to review transportation of commerci al fueled factory-fabricated modules in the near term.
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.19 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. Specifically, 10 CFR 71.41(c) allows for environmental and test conditions di fferent from those in 10 CFR 71.71 and 71.73 if the shippers controls provide safety of the shipment equivalent to that provided by meeting the regulations. Further, 10 CFR 71.41(d) p rovides for special package authorization if the application demonstrates that compliance w ith the regulations is impracticable and the safety s tandards established by the regul ations have been met through alternative means. Finally, an applicant may request an exemption as specified in 10 CFR 71.12. 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 package approval cannot eliminate the test but rather can reduce the severity of the test (e.g., the applicant can use a 20-foot drop
18 Type B packages contain a quantity of radioactive material greater than a Type A quantity. The NRC defines a Type A quantity of material in 10 CFR 71.4.
19 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).
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 22
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 only be approved through exemption.
After its experience with issuing the exemption for the Trojan reactor vessel (see ML20155E053 and SECY-98-0231, Authorization of the Trojan Reactor Vessel P ackage for One-Time Shipment for Disposal, dated October 2, 1998), the NRC noticed a need for a provision for a special package authorization for one-time shipment of large co mponents that do not meet the criteria for shipment as low specific activity packages or surf ace contaminated objects.20 In the 2002 proposed rulemaking ((67 FR 21390), Issue No. 12, Special Package Authorizations),
the NRC added the special package authorization option in 10 CF R 71.41(d) for limited circumstances involving large packages for which it is not prac tical to fabricate an authorized packaging. In particular, the NRC limited this alternate approv al method to, among other things, one-time shipments of large components for which compliance wi th the other provisions of these regulations [i.e., 10 CFR Part 71] is impracticable. The final rule (69 FR 3742, dated January 26, 2004) states that the special package authorizatio ns that will apply only in limited circumstances and only to one-time shipments of large component s. In order to meet the intent of the rule, a different application, and an NRC review resulti ng in a different approval would be needed for each shipment for each factory-fabricated module, li kely 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 categorical exclusion in 10 CFR 51.22(c)(13) for package designs for packages to be used for the transportation of licensed material s would not apply. In addition, each licensee making a shipment needs to request a separate exe mption, because an exemption cannot be made generical ly applicable to multiple lic ensees. Further, the DOTs regulations do not specifically authorize NRC-issued exemptions as a package approval; therefore, each licensee would need a DOT-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. 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. The NRC staff plans to continue to engage with indivi dual developers through pre-application activities and other stakeholders via the periodic advanced reactor stakeholder meetings.
20 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).
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 23
- 6. Storage of Fuel After Irradiation in a Power Reactor
Deployment Model Considerations
Fuel for factory-fabricated micro-reactors (and advanced reacto rs in general) may have much higher temperature limits than fuel historically used in light-water reactors, and there could be advanced reactor designs for which the spent fuel would not nee d to be water cooled immediately after withdrawal from the reactor, like existing zi rconium-clad, light-water reactor fuel. For advanced reactors, depending on the time duration bet ween withdrawal of the fuel from the reactor (or the final reactor shutdown) and placement into a dry storage facility, different regulations may apply to the storage of the reactor fuel or the fueled factory-fabricated module.21 Spent fuel must be cooled for at least 1 year prior to being s tored in an independent spent fuel storage installation (ISFSI) regulated under 10 CFR Part 72. Absent exemptions from the scope defined in 10 CFR 72.2(a) and other related requireme nts in 10 CFR Part 72, advanced reactor fuel that has not been cooled for at least 1 y ear would be required to be licensed under 10 CFR Part 50 or 52 for near-term storage as pa rt of the reactor facility, 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 includes criteria that the fuel has been withdrawn from a nucle ar power reactor following irradiation and has undergone at least one year's decay since b eing used as a source of energy in a power reactor. Therefore, the NRC could not license an ISF SI under 10 CFR Part 72 in which the spent fuel was decayed for less than a year.
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 year's 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
21 The NRCs definition of spent fuel in 10 CFR 71.4 is the same as that currently found in 10 CFR 72.3 for ISFSIs. The transportation and packaging requirements in 10 CFR Part 71 do not use the term spent fuel, such that the only location this term exists in the transportation regulations is in the definitions 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 be cooled for a year prior to transport, the NRC can approve package designs for shipment of fuel cooled for less than 1 year.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 24
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 radioactiv e 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 ar's 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 year's decay would be licensed under Part 50 rather than Part 72.
As stated in 10 CFR 50.1, the purpose of 10 CFR Part 50 is to provide for the licensing of production and utilization facilities, so the reference to li censed under 10 CFR Part 50 in the 1980 final rulemaking should be understood in that context. Sin ce that 1980 rule, the NRC has established 10 CFR Part 52 as an available path to license nucl ear 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 been cooled for at l east a year would be approved under 10 CFR Part 72. However, any near-term spent fuel storage of fuel that has been cooled for less than 1 year would be licensed under either 10 CFR Part 50 or 52, unless an exemption from the minimum cool time requirements of 1 year were granted to allow issuance of an ISFSI license under 10 CFR Part 72.
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.
- 7. Decommissioning Process and Decommission Funding Assurance
Deployment Model Considerations
Factory-fabricated micro-reactor deployment models might involv e transporting a factory-fabricated module away from the deployment site to a facility at a different location for decommissioning at the end of its life or for refurbishment and refueling before re-deployment. A decommissioning facility might be used to dismantle the factory -fabricated module to recover reusable parts and prepare the waste and spent fuel for transfe r, or it might also include an independent spent fuel storage facility. A refurbishment and re fueling facility might be used to defuel the factory-fabricated module, perform maintenance, refu el the module, and possibly operate the module for testing before re-deployment.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 25
In the following discussion, the NRC staff assumes that the dec ommissioning of factory-fabricated micro-reactors would involve independent decommissio ning of the factory-fabricated module and the deployment site. 22 This would require physically transferring the module from the deployment site to a decommissioning facility with the appr opriate license(s) to receive the module and its contents. The depl oyment site licensee 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 required if the facility were also to serve as the storage loca tion for spent fuel, power reactor-related Greater than Class C waste, and other radioactive mater ials associated with spent fuel storage. The facility would also need an operating license issu ed pursuant to 10 CFR Part 50 or a combined license issued pursuant to 10 CFR Part 52 to receive, possess, use the reactor in order to operate it for testing after refurbishment and refueli ng, 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 for testing.
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 71.59, whic h could potentially 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 for the purpose of decommissioning the deployment site. The deployment site licensee would be responsible for completing de commissioning of the deployment site consistent with applicable regulations of 10 CF R Part 20, Subpart E, as guided by NUREG-1757, Consolidated Decommissioning Guidance. The dep loyment site license for that reactor could be terminated upon meeting the decommissioni ng requirements for the
22 Factory-fabricated micro-reactors could also be decommissioned at the deployment site following the traditional approach used for large light-water reactors.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 26
remainder of the deployment site structures and systems unique to that reactor. If shared structures, systems, and components were to be used in connecti on with a replacement factory-fueled module, those structures, systems, and components import ant to safety or otherwise would need to be transferred to the new license for the replace ment module.
The deployment site licensee would need to establish decommissi oning funding assurance that considered the cost of removing the module from the site and de commissioning 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 20, subpart E. As r equired by 10 CFR 50.33(k)(1),
reactor applicants for an operating license or a combined licen se must provide a report as described in [10 CFR] 50.75, indicating how reasonable assuranc e will be provided that funds will be available to decommission the facility. For power reac tor licensees, reasonable assurance consists of a series of steps as provided in paragrap hs (b), (c), (e), and (f) of this section. The regulations at 10 CFR 50.75(c) establish the mini mum 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 1200 megawatts thermal be used i n calculating the minimum amounts, which is roughly 500 to 100 times the power level of f actory-fabricated micro-reactor designs. The formulas result in required funding assurance in e xcess of $75 million (January 1986 dollars) for each reactor, which is not compatible with de ployment models. Reliance on use of the minimum formula amount for decommissioning during op erations as reflected in 10 CFR 50.75(c) may need to be revisited as discussed in Near Ter m Strategy below.
In order to ensure adequate funding, a decommissioning cost est imate would need to consider and account for all activities and waste disposal costs associa ted with decommissioning the deployment site and decommissioning the factory-fabricated modu le at a decommissioning facility. It is possible that the decommissioning cost estimate for a module could be a predetermined estimate provided by a decommissioning facility l icensee or a refurbishment and refueling facility licensee authorized to acquire such a module (which could be the original fabricator of the module). It is also possible that the deploym ent site licensee would be the decommissioning facility licensee and a cost estimate would acc ount for the various component costs to dismantle and dispose of the module and store or trans fer the spent fuel. In either case, a preliminary decommissioning plan submitted with the deploymen t site license application would need to describe how the decommissioning funds would be a ccounted for between decommissioning the factory-fabricated module and the deploymen t site decommissioning activities. Later, decommissioning plans would be required to b e submitted in the form of a Post-Shutdown Decommissioning Activities Report, a site-specific Dec ommissioning Cost Estimate, a License Termination Plan, and a final status survey in accordan ce with 10 CFR 50.82 or 52.110.
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 use o f features to preclude criticality, then the NRC staff will consider this in its review of applicat ions for decommissioning and refurbishment and refueling facilities. The NRC staff intends t o also consider the approaches for licensing multi-module sites in SECY-11-0079 in relation to dec ommissioning and refurbishment and refueling facilities that may require 10 CFR Part 50 or 52 licenses.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 27
With respect to decommissioning funding assurance, the NRC staf f intends to consider specific exemptions under 10 CFR 50.12 from the requirements in 10 CFR 5 0.75(c) which were not established for non-light-water reactors. The NRC staff may consider site-specific decommissioning cost estimates in the review of applications fo r operating licenses and combined licenses for factory-fabricated micro-reactors. 23 The NRC staff notes that the use of site-specific decommissioning cost estimates is included in the proposed draft regulations in 10 CFR Part 53.
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.
- 8. 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. As discussed in SECY 0045, Population-Related Siting Considerations for Advanced Re actors (ML19143A194),
dated May 8, 2020, the attributes of advanced reactors, includi ng micro-reactors, are expected to provide a reduced 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 0045 - Population-Related Siting Considerations for Advanced Re actors (ML22194A885),
dated July 13, 2022, the Commissi on approved the NRC staffs proposal to revise the population-related siting guidance in Regulatory Guide (RG) 4.7, General Site Suitability Criteria for Nuclear Power Stations, Revision 3 (ML12188A053), issued March 2014, to provide technology-inclusive, risk-informed, and performance-based crit eria 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 guidance will state that, instead of locating a reactor in an area where the population density does not exceed 500 persons per square mile (ppsm) out to 20 miles from the reactor, an applicant can demon strate compliance with 10 CFR 100.21(h) by siting a nuclear reactor in a location where the p opulation density does not exceed 500 ppsm out to a distance equal to twice the distance at which a hypothetical individual could receive a calculated TEDE of 1 rem over a period of 1 month fro m the release of radionuclides following postulated accidents.
While the NRC is revising its population-related siting guidanc e to include alternate means of compliance with 10 CFR 100.21(h), the regulations in 10 CFR 100.21 remain unchanged. This includes the provision in 10 CFR 100.21(b), which requires that [t]he population center
23 The NRC staff described application of this approach to small modular nuclear reactors in SECY 0181, Decommissioning Funding Assurance for Small Modular Nuclear Reactors (ML112620358),
dated December 22, 2011.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 28
distance, as defined in § 100.3, must be at least one and one-t hird times the distance from the reactor to the outer boundary of the low population zone. In 1 0 CFR 100.3, population center distance is defined as the distance from the reactor to the n earest 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 consideration of population distribution rather than political boundaries. As such, current Commission policy and regulations would preclude siting a comme rcial power reactor, no matter the size or type of reactor, within a densely populated center. The allowable distance from the reactor to a densely populated center of approximately 25,000 r esidents 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 LPZ is required to be of such a size that an individual located on its outer boundary during the course of the postulated accident would not receive a radiation dose i n excess of 25 rem TEDE. The size of the LPZ depends on atmospheric dispersion characteristi cs and population characteristics of the site, as well as aspects of plant design. The NRC staff notes that 10 CFR 100.21 is not applicable to research and test reactors. However, testing reactors are subject to 10 CFR 100.11(a)(3) which requires a population center distance of at least one and one-third times the distance from the reactor to the outer boundary of th e low population zone. There are research reactors currently sited within population centers gre ater than 25,000 residents, and the NRC staff anticipates license applications for additional r esearch reactors to be sited in densely populated areas.
Deployment Model Considerations
Some micro-reactor license applicants may seek to site reactors at locations that are inconsistent with the current Commission policy and the regulat ions 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 its effort to revise RG 4.7 and wil l review license applications in accordance with current Commission policy that allows alternati ve population-related siting criteria but precludes siting a commercial power reactor, no ma tter 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.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 29
- 9. 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 appl ications 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 between several locations du ring their lifetime.
The various maritime applications envisioned by developers give rise to numerous potential legal, regulatory, and policy issues. For example, reactors loc ated in coastal waters may have different environmental considerations than land-based reactors. Such environmental 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 others. Deployment models might als o include scenarios where larger advanced reactors or light-water reactors are fabricated and potentially fueled in a factory before being transported to the maritime deployment site, which could potentially give rise to additional considerations beyond those examined for micro-react ors in this paper. Also, the current legal and regulatory framework may present challenges ( such as regulatory jurisdiction considerations, international licensing, and domestic licensing of operating reactors without fixed sites) for deployment models involving nuclear propulsion of commercial maritime vessels, including those where a U.S. flag vessel would operate in inter national or foreign waters, or a foreign flag vessel would operate in domestic waters.
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 DOE staff on maritime reactor act ivities through DOEs Maritime Nuclear Application Group.
- 10. 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
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 30
Administration (NASA) and US DOE are encouraging development of the technology, primarily for government projects. The NRC staff is not aware of any full y commercial ventures that plan to use micro-reactors for space applications, whether that be f or power generation 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 prior to launch activities.24 Upon initiation of launch activities, the Federal Aviation Adm inistrations (FAA) Office of Commercial Space Transportation (a part of the DOT) would ha ve authority.25 Terrestrial activities under NRCs regulatory jurisdiction would include li censing and oversight of the manufacture, construction, potential operation for testing or t echnology demonstration, transportation, and storage of utilization facilities intended to be deployed at extraterrestrial locations and in space. The NRCs authority under the AEA and 1 0 CFR Parts 50 and 52 for domestic licensing and regulation of utilization facilities doe s not extend to the operation of reactors outside the borders of 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 Government Age ncies on matters of regulatory jurisdiction, licensing, and safety of launches and application s of space nuclear systems as appropriate through ongoing interagency activities and the INSR B.
- 11. 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 Project Pele aims to develop and test a demonstration un it for a mobile micro-reactor for
24 Letter from Samuel J. Collins to Robert D'Ausilio (May 6,1998) (ML20013J130, nonpublic).
25 The Interagency Nuclear Safety Review Board (INSRB) issued a trial use playbook titled, Non-binding Guidance for INSRB and Its Counterparts, dated January 20, 2023, that includes, Appendix E: Defining a US Government Launch versus a Commercial Launch and DOT Authority. The appendix discusses the responsibilities and authorities of Federal Aviation Administrations Office of Commercial Space Transportation (a part of the DOT) as they relate to launches of space nuclear systems and states that, 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.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 31
military applications. NRC licensing is not required for this r eactor because it is authorized under AEA Section 91b., but DoD is seeking NRC approval of a transpor tation package under 10 CFR Part 71. The DoD engaged with multiple vendors who developed pr oposed designs for this program. Some micro-reactor developers have indicated that they may eventually 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). However, it would be difficult under the current regulatory framework to license commercial mobile micro-reactors for all potential operating sites. The NRCs approach to safety and environmental reviews presumes that a reactor will operate at a single site. Some of the specific technical requirements that would need to be satisfied in advance of deployment of a m obile micro-reactor are:
(1) The regulations in 10 CFR Part 100 establish approval requi rements for proposed sites for power and testing reactors subject to 10 CFR Part 50 or 52. The regulations at 10 CFR Part 100 specify that the requirements are for stati onary reactors, so the NRC staff would need to evaluate regulatory applicability (e.g., Could mobile reactors be considered stationary at each site of operation?) and how mobile micro-reactor applicants would meet the underlying purpose of the 10 CFR Part 100 requirements without knowing specific deployment sites ahead of time. 26
(2) NEPA requires Federal agencies to evaluate the impacts of p roposed federal actions on the human environment. The regulations in 10 CFR Part 51 for m the basis for the NRC's NEPA compliance and direct the NRC staff in how to perfor m environmental reviews. As a federal agency, the NRC must assess the environme ntal effects of proposed actions prior to making decisions. Therefore, movement of a reactor to a site that had not been previously permitted or licensed would n ot be allowed under the current regulatory framework. The NRC has never attempted t o perform an environmental review that would potentially cover deployment at innumerable, unspecified sites within the United States. The NRC staff would need to further evaluate the feasibility of performing such a review and possib le appropriate ways to meet the NEPA requirements, perhaps through bounding site param eters, etc.
Transportation requirements for f ueled micro-reactors are discu ssed in Section 5 of this enclosure. The transportation considerations for mobile micro-r eactors would be the same as those described in Section 5 for factory-fabricated modules loa ded with fuel that had been operated for some time (such as operational testing at a factor y or operation at a deployment site). Transportation package certifications are issued in acco rdance with the requirements in 10 CFR Part 71 and can be used to support an unlimited number of t ransportation events without additional licensing. However, if the Commission does not appro ve the NRC staffs proposal in this paper for the use of features to preclude criticality, the reactor would be considered to be in
26 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, applicants for particular sites perform site investigations and analyses to establish that their sites are appropriately bounded, and the NRC staff reviews this information for each site-specific application.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023) 32
operation when loaded with fuel and could not be transported be tween various sites for as-needed, where-needed power production.
Near Term Strategy
The NRC staff does not intend to take actions in the near term to address changes to the regulatory framework that may be needed to support mobile micro -reactor licensing.
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 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 ag encies and the need for development of a new regulatory framework.
Preliminary White Paper - Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, Enclosure (August 2023)