Fusion Power Plant Radiation Safety-by-Design Approach to Licensing

Pascal Dumont et al.

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1. Introduction

Abstract

We present the safety-by-design approach to nuclear licensing for a Stellarator Fusion Power Plant (S.F.P.P) that Type One Energy (T.1.E) utilizes in its engagement with nuclear regulators. T.1.E is pursuing licensing for Infinity Two, its First-Of-A-Kind (F.O.A.K) S.F.P.P in Tennessee, United States of America (U.S). The S.F.P.P presents a different radiation safety risk profile compared to N.P.P, which must be communicated clearly to regulators at every stage. Specifically, the U.S utilizes a byproduct material licensing framework for fusion machines as opposed to one that focuses on containment and control of fissile material.

To build a goal setting (objective-based) licensing framework and compliance program, the safety-by-design integration of physics, engineering, asset protection, and radiation safety constraints into structure, system, and component (S.S.C) requirements must be communicated to the regulator early in the Fusion Power Plant (F.P.P) design phase. In this paper, we first discuss the differences in safety profile and phenomenology that lead to a complete separation of F.P.P and fission Nuclear Power Plant (N.P.P) regulatory frameworks in the U.S and United Kingdom (U.K). The following sections outline the T.1.E approach to radiation safety analysis for calculating derived release limits for tritium, to identify S.S.C's radiation safety constraints early in the design and to address radiation safety during cyclical maintenance activities. We also address long-term considerations for developing codes, standards and a quality assurance regime for the F.P.P supply chain.

T.1.E S.F.P.P, Infinity Two is in the conceptual design phase for commercial deployment by the mid-2030s. The F.P.P presents a new paradigm for nuclear regulators experienced in N.P.P or mainstream byproduct materials licensing, and thus care must be taken to establish effective and rigorous, precedents for this new F.P.P industry.

Section 2 of this paper presents the radiation safety characteristics of fusion machine technologies that led the U.S Nuclear Regulatory Commission (N.R.C) in 2023 to adopt the byproduct material regulatory framework for licensing fusion machines including the need for a fusion-specific application guide to be developed.

The regulatory and licensing regimes for fusion machines in the U.S.A are rooted in the principles of radiation safety and recognize the relatively low radiological risk profile of emerging fusion technologies for power generation. figure 1 outlines some of the key characteristics of F.P.P's low radiological risk profile through comparison with N.P.P's. No construction permits or design certification are required in the U.S for licensing fusion machines. Only a license for the possession, production and use of by-product materials is required, supported by radiation safety analyses and evidence demonstrating that the control, containment and shielding of radiation and radioactive materials respect dose constraints and limits for the public and workers.

Figure 1 summary: A comparison table contrasting the hazards and risks associated with nuclear fission power plants and fusion power plants across three categories: fuel, radiation safety, and radioactive waste. Fission plants involve uranium-235, which produces highly radioactive waste and carries risks of core meltdown requiring emergency planning, while fusion plants utilize tritium and produce helium-4, with no potential for runaway chain reactions and waste that is manageable through recycling or low-level disposal. The takeaway is that fusion power plants present a fundamentally different and more manageable safety and waste profile compared to traditional fission reactors.

A key characteristic of fusion systems is that their principal radioactive material requiring control and containment is the circulating tritium fuel. In the event of a failure compromising the integrity of the primary radiological containment barrier, this gas can be promptly transferred to secondary controlled containment areas. By contrast, fission nuclear power plants utilize highly radioactive fuel which, if not adequately cooled during a reactor accident, presents a significant risk of severe core damage and environmental release.

The other source of mobilizable radioactive material consists of tritium-contaminated activated dust resulting from metal erosion localized within the vacuum vessel. Another favorable characteristic, this time for the Stellarator compared to Tokamaks, is that plasma stability greatly reduces the disturbances that increase and accelerate this erosion compared to Tokamaks.

This regulatory framework targets requirements calling for evidence of radiation safety that are very different from those in the framework for N.P.P's. The focus is on the identification of radiological inventories at the structures, systems and components (S.S.C) levels and their control, containment and shielding rather than on the safety classification of S.S.C's supporting the certification of the integrated design for N.P.P's.

However, these two distinct regimes have the same overarching objective, which is to ensure that dose limits and constraints for workers and the public are met and the principles of alara (As Low As Reasonably Achievable) respected.

To exploit its potential in terms of intrinsically safe design, a pragmatic and practical, technology-neutral approach must be used for the analysis and assessment of radiation safety, an approach that does not start from premises that are foreign to its own phenomenology, such as those established for existing nuclear power plants. This debate, which is currently closed in the U.S and U.K following decisions by the regulatory authorities, continues elsewhere in the world. Section 2 also presents a robust argument that can be used outside the U.S and U.K, based on the principles of'Defence-in-Depth' or DiD adapted to fusion machines.

This argument has also contributed significantly to the decisions to separate fusion from fission in the U.S and the U.K respective regulatory framework. Unlike fission reactors, fusion islands and their support systems, when they fail in their task of controlling and containing radiation and radioactive materials, cannot become uncontrollable in N.P.P-like accident progression scenarios where their fuel and core may be damaged (or even melt), resulting in radioactive releases that require the rapid implementation of off-site protective action measures for the public (sheltering in place, potassium iodine intake, evacuation, relocation, restrictions on locally produced food, etcetera). These releases from N.P.P's, which can be large-scale and uncontrolled (unfiltered) and occur very early in the accident process, may contain radioisotopes that are very harmful to the health of the population in the short term (e.g., Iodine-131) or long term (e.g., Cs-137).

Regulations addressing non-radiological environmental issues—such as water, air, and both hazardous and non-hazardous waste—are extensive and distinct from those governing radiological safety, which are the central focus of this paper. These regulations are prevalent across industries; however, their application and specific requirements can differ significantly between states and national jurisdictions. Section 3 presents the practical safety-by-design approach used by T.1.E with the Tennessee State regulator, focusing on the control and containment of radiation and radioactive materials in early radiological safety analyses to allow the early integration of radiological safety constraints into S.S.C design requirements. The primary objective of this paper, which focuses on the “safety by design” approach, is to explain how this approach ties into the design process within the goal-setting framework for licensing fusion machines under 10 C.F.R Part 30 (byproduct materials), rather than to provide a comprehensive summary of the final design results following implementation.

A key component of this safety-by-design approach is the systematic integration of radiological, engineering, and physical constraints into the formulation of design requirements across all systems and structural elements. These requirements are established using deterministic analysis tools such as the Master Logic Diagram (M.L.D) and Failure Mode Effect Analysis (F.M.E.A), initially at the functional level and subsequently throughout detailed design phases. The methodology also incorporates assessments of postulated initiating events, considering both internal and external site-specific hazards.

The primary goal is to ensure effective control and containment of radiation and radioactive materials, thereby achieving compliance with dose limits and regulatory constraints established for both the public and site personnel. This rigorous process begins in the early stages—covering pre-conceptual and conceptual design—and extends through the final design phase, consistently prioritizing radiation safety objectives.

Given that the regulatory framework for byproduct material is “goal-setting” and does not mandate a formal design certification process, applicants are not required to submit safety classifications for structures, systems, and components (S.S.C's); however, demonstrating safety with supporting evidence remains implicit and essential for license approval.

The paper provides practical examples illustrating how these constraints are incorporated into design requirements. It does not offer exhaustive details necessary for the plant's finalized design; instead, it emphasizes the pragmatic application of the safety-by-design framework within this goal-driven regulatory environment for fusion facilities.

Section 4 further expands on the concept of “safety by design”, discussing the iterative development and adoption of codes and standards specific to fusion technologies, as well as their maturation as operational experience is gained with First-Of-A-Kinds (F.O.A.K's) to Nth-Of-A-Kinds (N.O.A.K's) fusion power machines. Finally, Section 5 summarizes development of the radiation safety program for cyclical maintenance operations.

$ ^{*} $ Corresponding author at: Regulatory Affairs and Licensing, Type One Energy Group Canada. E-mail address: pascal.dumont@typeoneenergy.com (P. Dumont).

https://doi.org/10.1016/j.fusengdes.2026.115908

Received 18 February 2026; Received in revised form 30 April 2026; Accepted 22 June 2026

2. Regulatory landscape for licensing commercial fusion machines

Fusion energy developers face a variety of regulatory environments across the globe. In the U.S and U.K, fusion power plants have been placed within licensing regimes distinct from N.P.P's. However, in other jurisdictions, no clear and formal distinction has yet been established.

The Commissioners of the United States Nuclear Regulatory Commission (U.S N.R.C) unanimously decided in April 2023 to use the byproduct material regulatory regime as a basis for licensing fusion machines and rejected the N.R.C staff options for regulating fusion machines within the 10 C.F.R Part 50, 52 or 53 (reserved for Advanced Nuclear Reactors) frameworks.

This decision was underpinned by fundamental safety assumptions expressed in the U.S N.R.C staff's options paper “Options for regulating and licensing fusion systems” S.E.C.Y-23 to 0001.

The Commissioners' decision was rooted in public consultation outcomes found in S.E.C.Y-23 to 0001 conducted between 2020 and 2023 ^{} including the participation of D.O.E laboratories, international organizations like the U.K.A.E.A and I.A.E.A, startup fusion companies, various academics and researchers, etcetera S.E.C.Y-23 to 0001 outlines that fusion machines currently developed by companies represented by the Fusion Industry Association (F.I.A), which includes all companies (except those developing hybrid fission-fusion technologies) in the United States and Canada, have the following attractive characteristics from a radiation safety point of view:

- Do not involve any special nuclear material and, as such, the self-sustaining nuclear chain reaction that defines nuclear fission reactors is not possible in those fusion energy systems.

- Can benefit from access to low-activation materials, radiation shielding features and access to materials with no undesired levels of impurities to effectively eliminate Greater-Than-Class-C (G.T.C.C) waste so that most of the waste output will consist of low-level radioactive waste (L.L.W).

- Benefit from advances in plasma physics, auxiliary system technologies (e.g. superconducting magnets) and computing capabilities are enabling the design and operation of fusion energy systems that offer better plasma control. The latter is even more true for stellarators than for tokamaks, given the absence of induced current contributing to plasma disruption thus less dust production. The tritium stocks required for operation will be low, and it is these low stocks that are theoretically at risk of being mobilized in hypothetical failures or incident scenarios.

- Energy and radioactive material production from fusion reactions cease on varying timescales without any intervention in off-normal events or accident scenarios affecting radiological containment and active post shutdown cooling of the fusion containing radioactive material is not necessary to prevent a loss of radiological containment barrier (i.e., vessel breach). However, it should be noted that cooling will need to continue for some time after shutdown for asset protection reasons.

- Radionuclides present in the fusion systems, in processing or storage, or in activated materials, in any significant mobilizable amount are expected to result in low doses to workers and members of the public during credible accident scenarios (e.g., less than 1 rem (10 millisievert) effective dose equivalent to a person off site).

In June 2024, in line with N.R.C Commissioners' decision, the U.S Senate passed Bill S 870., which includes the Fusion Energy Act, legislation that codifies the permanent separation of regulations relating to fusion energy from those relating to nuclear fission and now defines what constitutes a fusion machine within the meaning of the Act.

The United Kingdom is the only other jurisdiction that has provided in its legislation for the separation of the regulatory licensing framework for N.P.P's and that for fusion technologies. The U.K Energy Act (Energy Bill) that was given Royal Assent on 26 October 2023 validates that fusion energy facilities have a significantly lower associated hazard than traditional nuclear (fission) sites and that nuclear site licensing would be disproportionate for fusion systems, as that regime was designed for sites with higher hazards than those presented by fusion systems. The Energy Bill confirms the existing regulatory approach to operational licensing of fusion research facilities can be utilized for commercial fusion energy systems; given that the radiological hazard of commercial fusion energy systems will be increased but not fundamentally different from current fusion research systems.

The byproduct material regulatory framework of 10-C.F.R-30 offers a risk-appropriate licensing path for minimizing dose to workers and members of the public. Agreement States (40 out of the 50 States) have delegated authorities from the U.S N.R.C to manage the licensing process under this byproduct materials framework. For fusion, the radiation safety focus is on the control, containment, and shielding of radioactive material and radiation present at the site, rather than on the performance and control of the machine S.S.C's. For example, in case of Loss-of-Cooling-Accident, Light Water Reactors fueled with cladded uranium oxide pellets must be provided with emergency core cooling system (E.C.C.S) designed with specific performance targets like peak cladding temperature and maximum cladding oxidation. For fusion machines, no such performance requirements exist in response to a loss of coolant flow as the consequences are not the same as explained above.

The expectation and regulatory requirement are objective calling for validation or evidence that, in the event of loss of coolant flow (say, in-vessel and for the first wall), there is no risk to the control and containment of radioactive materials and radiation leading to a dose in excess to the limits for workers and the public. Quality control during the manufacture and supply of these cooling systems and many other systems is advisable for the protection of property and the cost-effective operation of the machine but is not a requirement for the granting of a license under the 'by-product materials' framework.

Other countries have adopted the same approach of regulating fusion systems separately from N.P.P's but only for research and development systems if tritium, when present, does not exceed a certain threshold. Russia, for example, regulates R&D fusion machines outside the N.P.P regulatory framework if they do not use tritium or use tritium with a total inventory of less than 0.2 g. R&D machines containing greater than 0.2 g of tritium are licensed by the Russian Federal Service for Environmental, Industrial and Nuclear Supervision (Rostechnadzor) in accordance with Federal Law No 170-F.Z “On the Use of Atomic Energy” and the decree of the government of the Russian Federation “on the authorization of activities in the field of the use of atomic energy”. In France, any fusion system whether R&D or commercial authorized to contain greater than 27 g of tritium must obtain I.N.B licensing. These facilities fall into the same category as nuclear reactors, fuel cycle facilities, storage facilities, the most powerful electron and ion accelerators and all facilities using significant quantities of radionuclides, in accordance with Article 50 593 to 2. of the Environment Code.

For T.1.E's S.F.P.P and many other fusion machines, the radiological hazards to be controlled and contained are (i) prompt plasma sourced neutrons and x-rays, (2) parasitic losses of tritium (beta particles) and (3) activated system materials (photons). Radiation protection design features and operating procedures will be necessary to control human exposure to radioactive materials and minimize exposure to plasma-sourced radiation.

2.1. Goal-setting licensing regimes: facilitating innovation

Instead of mandating specific engineering requirements, 10 C.R.F Part 30 (byproduct materials regulations) and 10 C.F.R Part 20 (radiation protection standards) establish fundamental radiation safety objectives. Under this approach, a license is approved if an applicant demonstrates that their radiological barriers, related equipment and systems, and procedural controls effectively protect the public, workers, and the environment from ionizing radiation risks.

As outlined in the 2022 U.K.A.E.A report summarizing public consultations on regulatory frameworks for fusion technologies in the United Kingdom, a goal-setting regime emphasizes achieving defined objectives rather than rigid adherence to predetermined performance standards systems. This methodology acknowledges that those who assess risks while performing radiation safety analyses informing design requirements are most equipped to manage them. By setting overarching goals—rather than imposing prescriptive requirements that may not be universally applicable to emerging technologies with limited operational experiences—license applicants gain flexibility to determine the most suitable solutions for their specific radiation safety risks. This fosters innovation and facilitates the advancement of novel fusion commercial technologies within a comprehensive, goal-oriented framework.

Within this context, enabling regulators such as the Health and Safety Executive (H.S.E) in the United Kingdom and Agreement States in the United States serve a vital function in supporting and overseeing innovative commercial fusion technologies. Their active participation ensures that regulatory practices remain adaptive to technological progress, maintaining effective oversight while encouraging ongoing advancement.

2.2. Differentiating fission and fusion power plants: implications for fusion licensing regime

With such small amounts of tritium as seen in the previous section (0.2 g in Russia and 27 g in France), to create a bifurcation in the regulations applicable to fusion machines, these regulatory frameworks appear to pay little focus to the completely different operational dynamics of fusion compared to fission technologies. Fission plants produce tritium and manage it as an undesirable contaminant, whereas for fusion it is a valuable fuel that cannot be lost without jeopardizing the economic viability of operations. This logical connection is also unclear, since these small quantities of tritium do not compare in risk profile with the significantly larger quantities (tons) of uranium used in fission plants. As demonstrated later in this paper, the economics of tritium fuel and tritium radiation safety issues are closely linked.

To be economically viable, with tritium trading at around $30,000 U.S.D per gram, parasitic losses must be contained within multiple radiological containment barriers, reclaimed, recovered and reintegrated in the fuel cycle. By targeting through design, the fuel savings necessary for the sustainability of the fusion machine, radiological safety issues are implicitly considered and addressed.

The radiological containment barriers referred to throughout this document comprise several functional hierarchical levels and are essential design features for the control and containment of radiation and radioactive materials. The exact design and systems used for these barriers vary from one technology to another and are currently being evaluated for the T.1.E S.F.P.P. General descriptions of radiological containment barriers commonly presented in the literature show the vacuum vessel (and some systems attached to the vacuum vessel such as the fuel cycle systems) as constituting the primary containment, and the fusion machine building envelope as the ultimate radiological containment barrier.

An intermediary radiological containment barrier encompassing the primary one is presented in [15] and designed to cope with local overpressures due to challenging normal operations or abnormal events. It includes all the S.S.C's containing activated materials and/or radioactive materials, and particularly the expansion volume (connected to the vacuum vessel, cryostat, cooling pipe gallery, steam generator room, blanket purge gas loop and other volumes at risk of over pressurization) to limit local design pressures to reasonable values.

Whether at the European Commission, the International Atomic Energy Agency, the Clean Air Task Force or within alliances of countries such as the Agile Nations Initiative, efforts are underway to establish guiding principles to ensure harmonized regulatory practices to facilitate the rise of fusion technologies as a clean, safe, and fuel-abundant commercial energy source around the globe. These initiatives raise questions and reveal sometimes contradictory and competing viewpoints on the need to adapt existing regulatory frameworks for fission-based nuclear power plants to the realities of fusion technologies. As fusion technologies will continue to be compared with those of nuclear fission reactors in the regulatory framework across the globe, it is reasonable and productive to use the same concepts and definitions of radiological safety to fuel fair and productive debates.

It would be a great loss if member states were to deprive themselves of such a clean and promising source of energy by basing their decisions on false assertions regarding radiological hazards and risks. Nations such as Australia, New Zealand, Thailand, and Singapore, without a fleet of fission nuclear power plants but with regulated medical and/or industrial nuclear applications, are well placed to plan for the adoption of fusion systems as a source of clean energy production.

2.2.1. The principles of 'Defence-in-Depth' (DiD) as an anchor point to explain the phenomenological differences

T.1.E's S.F.P.P behavior or phenomenology can be analyzed through the lenses of the DiD principles which have been widely adopted internationally. This practice will help to clarify fundamental differences with N.P.P radiation safety hazards and risks and the resulting design rules supported by prescriptive safety analysis and assessment requirements. Levels 1 and 2 of the DiD principles, as illustrated in figure 2, are relevant for Fusion Power Plants (F.P.P's) and T.1.E S.F.P.P design efforts have followed and continue to follow level 1 and 2 DiD objectives from the earliest stage of the pre-conceptual design.

Figure 2 summary: The defense-in-depth levels for Nuclear Power Plants (NPPs) and Fission Product Plants (FPPs) differ primarily in the number of levels and the strategies employed. NPPs use five levels of defense, whereas FPPs use only three. Both plant types utilize Prevention strategies for levels one and two during normal operation and anticipated operational occurrences. However, from level three onward, the focus shifts to Mitigation. For FPPs, level three handles design-basis incidents or events using engineered safety systems and on-site incident operating procedures. NPPs extend mitigation through level four for design-basis accidents and level five for beyond design-basis accidents and post-accident recovery. Level five for NPPs includes severe accident management guidelines to contain radiological material and mitigate consequences, while the means of control for the final stage of NPP defense involves off-site protective actions and monitoring.

Process systems (starting at level 1 with normal operation states) and detection and control systems (for covering level 2, that is, anticipated operating occurrences) are the means of control which will be supported by operating procedures and programs. Level 2 covers challenging normal operations (failures and/or tran-zee-unt with no critical impacts on the fluctuation of S.S.C's relevant to radiation safety outside safety margins) and abnormal operations (e.g., loss of external power mitigated by emergency power classes with potential for S.S.C's performance and reliability fluctuating outside safety margins). Design analysis (level 1) and deterministic safety analyses (level 2) using the Master Logical Diagram (M.L.D) approach at pre-conceptual design phase and Functional Failure Mode Effect Analysis (F.F.M.E.A) at the conceptual design phase will be employed. This will enable the identification of regulatory constraints for each safety relevant S.S.C's (relevant to the control and containment of radiation and radioactive material) to meet dose constraints and limits in all modes of operations while aiming for alara.

It is from level 3 onwards that the similarities between fission and fusion cease for this DiD model, given the phenomenological differences that dictate a sequence without progression towards accident states (Design-Basis Accident and Beyond Design-Basis Accident) for fusion machines. figure 2 highlights a level 3 specific to fusion that has nothing in common with the third level of DiD for N.P.P's and the absence of levels 4 and 5. The third level for fusion machines is a return to normal operation after a machine shutdown due to a failure to detect and/or control an abnormal or challenging operating situation. This could be a machine shutdown due, for example, to a broken cooling pipe inside the vacuum vessel (first wall or blanket or divertor cooling pipes for example), which would have led to an increase in the pressure of the vacuum vessel and the need to move the D-T fuel to the expansion volume (this specific case will be covered as an example in the next section).

This is therefore a phase of stabilization and repair (incident or failure mitigation) with a view to resuming operations. The N.P.P level 3 DiD is a design basis postulated accident involving accident progression leading to accident conditions for which a facility is designed in accordance with established design criteria and conservative methodology, and for which releases of radioactive material are kept within acceptable limits in terms of emergency dose exposure situation. Such an N.P.P situation requires the activation of off-site and on-site emergency measures as well as the use of safety systems to prevent progression towards a beyond-design-basis-accident (level 4) involving severe accident conditions calling for the implementation of on-site and off-site protective actions which represents the level 5 DiD for N.P.P's.

The differences and phenomenological advantages of fusion from a radiation safety perspective that justify an adapted DiD model are based on three (3) general aspects:

(i) The stored energy in the fuel in the core of a fission power plant is 5 to 6 orders of magnitude larger than that in a fusion power plant (fractions of a gram in fusion, tens to hundreds of kg in fission). It is also several orders of magnitude larger than the energy stored in the magnetic field of the stellarator magnets, or any other part of the system. This makes the safety by design solutions against fault scenarios in a fusion power plant orders of magnitude smaller than those for fission.

(2) As there is no fissile material present and criticality (a self-sustaining neutron chain reaction) is not possible. The power and radiological inventories present when the machine is shut down whether voluntarily or not, can't increase and sustain any incident progression.

(3) No releases of radionuclides are expected that could result in a dose equivalent to the public above the regulated dose threshold of 1 rem (10 millisievert): this dose represents the regulatory threshold calling for the requirement for off-site emergency preparedness and response plan in the U.S. T.1.E is designing its tritium fuel cycle systems, and peripheral systems impacting the tritium fuel economy with the aim of reducing the operating inventory at risk of mobilization within levels that cannot exceed the public dose limit of 1 millisievert.

While nuclear power plants contain tens of tons of uranium fuel within the reactor core, fusion systems operate with only a very small quantity of fuel present in the plasma at any given time. The actively burning plasma, the region where fusion reactions occur, contains on the order of milligrams to at most a few grams of deuterium-tritium fuel.

3. Radiation safety by design – methods & analyses

The 'Safety-by-Design' approach requires iterative considerations from the earliest stages of design to integrate radiation safety, physics and engineering (asset protection) constraints into the requirements for S.S.C and their interfaces. Table 1 presents the main topics of discussion associated with the four (4) parts Title 10 of the Code of Federal Regulations that play the most significant role in identifying radiation safety constraints and that drive discussions with the Tennessee regulatory authority prior to licensing, starting at the pre-conceptual design phase.

It should be noted that the high energy photons emitted by the plasma are negligible from a radiation protection standpoint compared to the neutron flux and will primarily be considered for their contribution to the heating of the first wall.

Although all these topics are discussed with the Tennessee regulatory authorities to whom authority has been delegated from the U.S. Nuclear Regulatory Commission (N.R.C) to regulate byproduct materials within their borders, this section addresses the ones related to the following two aspects:

(i) The performance and reliability of tritium fuel cycle systems: Such systems dictate both the operational tritium inventory required and, therefore, the quantities of tritium at risk of being mobilized in work areas. A controlled atmospheric release of tritium could also be necessary to maintain the occupational dose within the required limits when collection of tritium parasitic losses is not practical nor possible. These atmospheric emissions will be subject to regulatory limits established by calculating the derived release limit (D.R.L), which informs the reference design of the S.F.P.P.

(2) The control, shielding and containment of radiation and radioactive materials must be demonstrated to the regulatory body in the design, before starting construction, for all S.S.C's playing a direct or indirect role in radiation safety to maintain dose constraints and limits for workers and the public (S.S.C's relevant to radiation safety). It should be noted here that this is not an exercise leading to design certification according to design acceptance criteria, but rather information that allows for validation.

10 C.F.R Part 30 (Byproduct Materials): Sets the conditions for possession and use of certain types of radioactive materials

- Inventories of radionuclides at S.S.C level

- Maintenance plan & radsafety aspects

- Focus on S.S.C's involved with control and containment of radiation & radioactive materials

- H.T/H.T.O atmospheric dispersion and exposure pathways leading to dose projections for critical groups

10 C.F.R Part 20 (Radiation safety): Standards for protecting people from ionizing radiation

• Off-site dose constraints set at 1/10 of public dose limits for H.3 releases

• Fuel/Tritium cycle systems and H.V.A.C design performance critical for internal dose management

• Plasma source term (neutrons), shut-down dose rate (activation) and shielding optimization

that the applicant meets regulatory expectations at this stage of the project from an objective point of view. There are therefore no performance requirements for S.S.C's, as is the case with 10 C.F.R Part 50, 52 and 53 (see previous example above related to §50.46 for acceptance criteria for emergency core cooling systems for light-water nuclear power reactors).

3.1. Safeguard by design for fusion commercial machines

Although this paper does not focus in detail on the “safeguard-by-design” aspects of fusion machines, it is important to recognize key points outlined in the I.A.E.A Fusion Key Elements document. For fusion designs that neither process, use, nor produce special nuclear material (S.N.M), the I.A.E.A does not implement or recommend safeguards measures. Commercial fusion machines, as represented by the Fusion Industry Association, are not among those that would utilize S.N.M.

Fusion technologies differ fundamentally from nuclear power plants (N.P.P's) in that they do not use or produce special nuclear materials required for weapons-grade nuclear material production. If any fertile or fissile material, such as U-238 or U-233, were covertly introduced into a fusion power plant, monitoring equipment like gamma ray detectors would promptly detect its presence—typically within minutes. In contrast, N.P.P's contain U-238 and U-233 as standard components, so the detection of their gamma rays does not necessarily indicate an anomaly.

One major advantage of fusion technology compared to fission is that, once authorized operations begin, no fissile material is present in the reactor. Furthermore, fusion systems are not designed to enable the transmutation of S.N.M into weapons-grade nuclear materials.

Introducing small amounts of fertile particles illicitly into the breeding blanket and its cooling system would require extensive modifications over several months. This process would also demand considerable resources and time to develop infrastructure capable of processing and manufacturing the transmuted nuclear material into weapons-grade forms.

3.2. Evaluating the maximum amount of tritium that can be mobilized outside the process boundary (first radiological containment barrier)

Parasitic losses of tritium fuel, that are unavoidable, can contribute to worker exposure, in addition to photons, if not reclaimed and recovered by air detritiation systems for example. The amount of tritium 10 C.F.R Part 37 (Physical Protection): Requirements for physical security of certain categories & quantities of radioactive materials

• Metal impurities (e.g., cobalt in steel) to be avoided

- List is fission-driven and need to be updated for fusion

• Applies mainly for refurbishment during cyclical maintenance & decommissioning activities

- Early engagement and data sharing with Agreement States and U.S N.R.C

10 C.F.R Part 61 (Radwaste management): Licensing requirements for land disposal of radwaste

• Early evidences on novel waste stream/radionuclides key for regulatory improvement • Technology maturation for recycling approaches is promising and essential • The recycling of L.L.W is a global challenge affecting the A.N.R's and N.P.P industries, requiring synergistic efforts.

required in circulation (operating inventory) is thus an important performance indicator in that if this amount is kept to a strict minimum, as required by alara and economic incentive, the health risks to workers and the public is thus well mitigated. This also applies to off-site public health, as atmospheric releases aimed at reducing the contribution of tritium to the total dose received by workers in cases where escaped tritium capture is impossible due to system failures will be alara. It is also important to note that this same operational inventory is the one that can be mobilized, in part or entirely, in the event of credible large-scale external or internal loads (e.g. fire, explosion, earthquake, deliberate malicious acts, etcetera). If this quantity is relatively small, the effects on human health will remain well below the dose limits requiring off-site emergency preparedness and response, set at 10 millisievert (1 rem).

T.1.E's Infinity Two S.F.P.P is pursuing a distinctive integration of advanced technologies with mature and proven systems to support deuterium-tritium (D.T) operations. The tritium fuel cycle architecture, comprising the tritium plant, fueling systems, blanket systems, and vacuum systems is under development with inherent safety principles and optimized to minimize radiological risk. This approach aligns with best practices in fusion system design and is consistent with the defense-in-depth philosophy for radiological protection.

This commitment is exemplified in our recent publication, which presents a detailed residence time modeling and optimization study. The study demonstrates how Infinity Two can significantly reduce both the start-up tritium requirement and the operational tritium inventory, as summarized in Table 2.

: Table 2 summary: Tritium inventory metrics for the Infinity Two pre-concept design show that the Tritium-Lean scenario significantly reduces inventory compared to the Baseline. In the Tritium-Lean case, start-up inventory drops from 3.701 kilograms to 0.843 kilograms, and operational inventory decreases from 0.673 kilograms to 0.233 kilograms. This reduction is accompanied by a higher Tritium Breeding Ratio, which increases from 1.040 in the Baseline to 1.08 in the Tritium-Lean scenario.

Where the following definitions are noted:

3.2.1. Start-up tritium inventory

The start-up tritium inventory is defined as the quantity of tritium required to initiate plasma operations and commission the tritium fuel cycle prior to achieving sustained, self-sufficient breeding. This inventory supports initial plasma fueling, conditioning of in-vessel and fuel-cycle systems, and establishment of steady-state processing flows. During plant operation, this tritium is progressively redistributed into system components and materials and is not required to remain fully available as circulating fuel once steady state conditions are achieved.

3.2.2. Operational tritium inventory

The operational tritium inventory is defined as the amount of tritium that must be actively present and mobile within the fuel-cycle systems to sustain normal steady-state operation. This includes tritium circulating through plasma fueling systems, short-term storage, processing units, and blanket extraction systems, and represents the minimum inventory necessary to maintain continuous operation without reliance on additional external supply.

3.2.3. Relationship between start-up and operational inventories

The difference between the start-up inventory and the operational inventory represents tritium that becomes retained within plant systems and structures during operation, including storage media, processing equipment, piping volumes, and material interfaces. This retained tritium is not assumed to be simultaneously available for plasma fueling and is not equivalent to the actively circulating operational inventory.

3.2.4. Baseline and tritium-lean scenarios

The Baseline scenario reflects a conservative design approach that maintains larger start-up and operational inventories to provide operational margin and system buffering. In contrast, the Tritium-Lean scenario minimizes both start-up and operational inventories through reduced buffering and tighter coupling between breeding, processing, and fueling systems, thereby lowering the total tritium present within the plant at any given time while maintaining fuel self-sufficiency through a correspondingly lower required tritium breeding ratio.

The table and accompanying discussion present values derived directly from the analysis reported by Clark et al.. That work explicitly defines the scenarios considered and the corresponding inventory requirements, including a baseline start-up inventory representative of tritium fuel-cycle technologies currently available, as well as a reduced start-up inventory scenario enabled by significant advancements in tritium handling and processing technologies. The definitions of the baseline, tritium-lean, and operational inventory concepts are clearly articulated in the referenced study and provide appropriate support for the material presented here. For clarity and completeness, these definitions have now been explicitly incorporated into the present manuscript, as provided in the additional explanatory text above.

The design incorporates Direct Internal Recycling (D.I.R) technology, which redirects unburned fuel from the exhaust directly back into the fueling system. This closed-loop configuration reduces the need for large tritium buffers and minimizes the load on purification systems, while maintaining a continuous and efficient fuel cycle. Minor adjustments to this loop allow for impurity removal and tritium replenishment.

Additionally, an online tritium extraction system is planned to maintain residual tritium concentrations in the blanket at alara levels. This strategy avoids the accumulation of large inventories for batch detritiation, thereby reducing both operational dose and residual risk within the bio-shielded environment.

All tritium-handling systems, including vessels, piping, valves, and pumps, will be constructed from materials resistant to tritium-induced degradation and designed to meet stringent mechanical integrity, vacuum, and pressure containment standards. However, recognizing that the complete elimination of tritium leakage is impractical, the design employs a multi-barrier defense-in-depth approach.

This includes:

• Primary system containment.

- Secondary confinement (e.g., gloveboxes, ventilated and isolated rooms).

• Tertiary barriers (e.g., building envelope).

- Dynamic barriers such as Gas Detritiation Systems G.D.S.1 and G.D.S.2 and Air Detritation Systems (A.D.S).

These systems maintain a cascade of sub-atmospheric pressures to ensure airflow from clean to contaminated zones. In normal operation, G.D.S 2 detritiates effluents and maintains glovebox pressure. In the event of a leak, G.D.S 1 and A.D.S isolate and decontaminate affected areas, ensuring safe discharge to the atmosphere through a high-efficiency stack designed to limit public exposure below regulatory thresholds as illustrated in figure 3.

.Figure 3 summary: A schematic diagram illustrating a hierarchical system of enclosures for tritium confinement within a nuclear building. The system flows from the innermost Primary Equipment, enclosed by a Vacuum Jacket and a Glove Box, to a secondary Room enclosure, and finally the tertiary Nuclear Building boundary, with Gas Detritiation Systems (GDS1, GDS2) and an Air Detritiation System (ADS) managing atmospheric purification. The purpose of this nested structure is to provide multiple redundant barriers to prevent the release of tritium into the environment.

In the tritium-lean scenario, the inner fuel cycle—comprising the vacuum pumping chain, D.I.R loop, fueling system, and isotope separation unit—holds the largest operational tritium inventory, approximately 0.667 kilograms. This underscores the importance of robust confinement and detritiation strategies in these subsystems.

There is currently a trend in the deliberations of various jurisdictions regarding the best regulatory and licensing framework for fusion machines which focuses on tritium inventories as a key criterion for determining whether fusion systems should be regulated under a fission-based regime. This would be in-line with fusion regulations in France and Russia, as noted previously.

Concerns have been raised about the risks associated with tritium management for fusion machines based on the constraints encountered in managing tritium from N.P.P's operations. However, several important distinctions highlight why the fusion fuel cycle and associated tritium inventories should not be compared to the challenges posed by tritium in N.P.P's.

- State of tritium in the fusion cycle: In the fusion tritium fuel cycle, tritium is in gaseous form. This is fundamentally different from the N.P.P's context, where tritium is primarily found as tritiated heavy water (D.T.O) and tritiated water (H.T.O), a byproduct of neutron activation in heavy water moderators. The radiological behavior, containment strategies, and associated risks differ significantly between these two states.

- Purpose and value of tritium in fusion: Tritium is a valuable fuel - not a contaminant. It is carefully managed, recycled, and accounted for within closed-loop systems in gaseous form with avoidance of capture by water. In contrast, the N.P.P tritium removal facilities are designed to remove and dispose of tritium from heavy water to reduce radiological hazards. This difference in intent drives very different operational and safety paradigms.

- This situation is fundamentally different from that in most nuclear power plant (N.P.P) contexts, where a significant fraction of the tritium inventory exists as tritiated water, principally tritiated heavy water (D.T.O) and tritiated light water (H.T.O), produced continuously through neutron activation processes in heavy-water moderators and associated coolant systems.

In fusion systems, tritium is primarily managed within the fuel cycle in elemental or molecular form; however, depending on the reactor concept and materials selection, a portion of the tritium inventory may permeate into coolant systems or undergo isotopic exchange reactions resulting in the formation of tritiated water (H.T.O). Additional H.T.O may arise within auxiliary systems such as glovebox atmospheres, heating, ventilation and air-conditioning (H.V.A.C) systems, or following confinement breaches. In some fuel-cycle configurations, tritium converted to H.T.O is routed through dedicated water detritiation systems and subsequently reprocessed or reintroduced into the fuel cycle. Unlike heavy-water fission reactors, where tritiated water is generated continuously as an inherent byproduct of reactor operation, the formation of H.T.O in fusion facilities is generally limited in magnitude, system-dependent, and managed through controlled processing pathways.

- Radiological and safety profiles: The radiological profile of gaseous tritium is distinct from that of H.T.O. Gaseous tritium poses flammability and permeation risks, which are well-understood and have been managed in industrial settings for decades. H.T.O, on the other hand, presents challenges related to biological uptake and environmental dispersion, necessitating different mitigation strategies.

Routine tritium releases, when necessary, are not expected to approach regulatory thresholds related to dose constraints or dose associated with even more restrictive administrative limits. If such a release is necessary due to a failure (abnormal operating condition) of the air detritiation systems that would result in the dose threshold for workers being reached, such a release would be marginal in terms of the dose to the public (well below the threshold of 0.1 millisievert for one year) in all scenarios. Preliminary simulations performed with the HotSpot health physics code show that such a release (the entire mobilizable inventory of 700 g) in a context of high external and/or internal energy load event (hazard) would result in a hypothetical dose well below the 10 millisievert limit that call for the requirement of an off-site emergency preparedness and response plan.

3.3. Calculating Tritium Derived Release Limits (D.R.L's) for Planned Routine Releases

D.R.L calculations for tritium are performed at two levels. First, to inform the design reference level, a design that must be compatible with as many potential site conditions as possible worldwide. The D.R.L calculation for the reference design considers a wide range of exposure pathways for the most critical person in terms of dose projection.

Next, the same calculation must be performed for the specific characteristics of the sites desired for the construction of the S.F.P.P to validate that the design reference level corresponds to the site conditions in terms of exposure pathways for the most critical individual. Site-specific D.R.L's are also those that are enforced by the licensing conditions and therefore agreed with the regulatory authorities.

T.1.E is using the Canadian Standard Association (C.S.A) N288.1 guideline and model for calculating the D.R.L's for meeting regulatory public dose constraints and limits for tritium airborne release for normal operation.

3.3.1. Background on the calculation of derived release limits using C.S.A-N288.1

In Canada and other candu operating countries, the control of tritium releases to the environment is particularly important as the Canadian-designed candu Canada Deuterium Uranium) reactors produce more tritium than most other types of reactors. Unlike Light Water Reactors, these reactors use heavy water in their moderator and heat transport systems (heavy water is normal water in which the hydrogen atoms have been replaced by deuterium atoms) which produce tritium because of neutron capture in deuterium.

Routine releases from candu Power Plant tritium into the environment have been carefully monitored and regulated for decades. Tritium concentrations in the environment (in air, water, vegetation, animals and milk) are measured to estimate the annual dose to the public living near nuclear facilities. This information is used to confirm that the impact of the releases is far below the annual public dose limit of 1 millisievert (100 mrem), as established in the Canadian Nuclear Safety Commission's (C.N.S.C) Radiation Protection Regulations.

Tritium air concentrations in the vicinity of nuclear facilities varied between 0.38 becquerels/m cubed 1.03E-05 μCi/m cubed) and 35.66 becquerels/m cubed 9.64E-04 μCi/m cubed). The corresponding dose due to exposure to tritium for people living near nuclear generating stations varied from 0.00045 millisievert/year (0.045 mrem/year) to 0.00236 millisievert/year (0.236 mrem/year). Tritium related exposure to members of the public around processing facilities were also very low (0.00001 to 0.0145 millisievert/year). All doses are well below the regulatory annual dose limit for a member of the public.

The operational environmental data from the candu facilities are relevant to provide context toward T.1.E's S.F.P.P tritium release limits for the conceptual design phase. While different, this historical data can be used to inform optimal siting options and performance and confidence in systems and processes involved in the tritium fuel cycle.

Derived release limits (D.R.L's) are administrative limits for demonstrating that routine facility releases meet the regulatory public dose limit of 1 millisievert (100 mrem) for site-wide releases and the dose constraint of 0.1 millisievert (10 mrem) for atmospheric gaseous emission of tritium. These estimates are calculated by license applicants and once accepted by the regulator, become a legally binding license compliance. requirement. D.R.L's are calculated using an environmental transfer model, such as one depicted for tritium releases, in figure 4. Details related to D.R.L environmental modelling can be found in C.S.A Standard N288.1 ^{} .

><?=$ Figure 4 summary: A flow diagram of a Tritium Dynamic Radio-Logical (DRL) model showing the movement of tritium between different environmental compartments. Tritium moves from a Source through the atmosphere and surface/groundwater into various biological and geological sinks, including soil, forage, animal produce, aquatic animals, aquatic plants, and sediment. All these compartments eventually contribute to the total human dose. The model illustrates how tritium cycles through the ecosystem before ultimately reaching humans.

The Canadian Standards Association (C.S.A) guideline for calculating derived release limits for radioactive material (which include tritium-specific parameters) in airborne and liquid effluents is an internationally recognized standard. It is consistent with and makes direct reference to data and approaches in reports, publications, standards and guidelines of the:

• International Atomic Energy Agency (I.A.E.A):

• the International Commission on Radiological Protection;

- the U.S National Council of Radiation Protection and Measurement (U.S N.C.R.P;

- U.K Ministry of State for Agriculture, Fisheries and Food (U.K M.A.F.F;

• U.S Environmental Protection Agency (U.S E.P.A;

• U.S Nuclear Regulatory Commission (U.S N.R.C):

- as well as various scientific articles and R&D organizations.

To effectively determine a D.R.L, the model works from source to receptor, solving the contributions to each compartment in each pathway. The process involves basic concepts (dose limits and representative persons), an environmental transfer model (pathways for representative persons exposure to radionuclides) and development of transfer parameter values for the model. The transfer parameters are developed for a radionuclide release to the atmosphere and for a release to surface water. The model may be used to calculate the radionuclide activity in each environmental compartment based on a unit release in becquerels/s (Ci/s) and the associated human dose in millisievert/a (mrem/a). The D.R.L is then calculated as the applicable dose limit (100 mrem/yr or 1 millisievert/y to a number of the public) divided by the dose per unit release calculated using the environmental transfer model (millisievert/yr per becquerels/s). The result is the D.R.L in becquerels/s, conventionally presented as a value in becquerels/yr (Ci/yr).

To solve the details of the model, representative values for physical dispersion, bioaccumulation parameters, and reasonable maximum values for human intakes or external exposure factors are required, such that a reasonable maximum dose per unit release is obtained. A conservative approach provides assurance that dose per unit release will not be underestimated, and that D.R.L's will adequately protect human health. Although it is possible to utilize default or generic transfer parameters to estimate a D.R.L, use of site-specific values is most desirable if available.

As seen in figure 4, the source transfers initially to air and water pathways. While both models are complex to solve, airborne dispersion is a very well-developed technical field. In addition, airborne release is affected by meteorological conditions that direct the release by parameters such as windspeed and direction, temperature, etcetera By comparison, waterborne release generally occurs through directed pathways from the source location to a given surface or ground water receptor, which will have specific flows.

For example, an airborne release will be governed by the wind rose, with the possibility of dispersion in all directions, whereas a waterborne release into a river flowing east to west will bound that release to this direction (to a first approximation). While routine airborne releases are from a facility stack, routine waterborne releases may originate from areas such as condenser cooling water (typical for nuclear power plants).

Transfer factors, required to solve for the tritium content in the compartments, must be obtained via available guidance or from locally accepted values.

3.3.2. D.R.L's Calculation for the S.F.P.P's Design Reference Level

International Commission on Radiological Protection guidance [29] recommends limiting the public annual radiation dose limit 1 milli-sievert (millisievert) from artificial sources, excluding medical procedures. The derived tritium D.R.L's for meeting regulatory requirements are defined for the public dose limit (1 millisievert) and dose constraint for air emission of tritium (0.1 millisievert). The dose constraint is used to implement the alara [as low as is reasonably achievable] requirements of 10 C.F.R 20 §20.1101 (b), and notwithstanding the requirements in §20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon-222 and its daughters, shall be established by licensees other than those subject to 10 C.F.R 50 §50.34a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 millisievert) per year from these emissions. This requirement is enforced in Tennessee.

D.R.L's were calculated using “bounding site scenarios” where all potential parameters (i.e., credible radiation exposure pathways) are considered effective and the atmospheric conditions and effective point of release of tritium are set realistically (assumed 60 m release height and using local meteorological data for population and land within the dominant wind sector; the selection of 60 m release height is based upon a typical stack height used for fission power plants for the purpose of the D.R.L calculations here, will be adjusted to real stack height based on engineering design, and is typically based upon some multiplying factor applied to the height of nearby structures).

This allows T.1.E to design the performance and reliability of the Tritium Fuel Cycle (T.F.C) S.S.C's and fit in a vast array of site conditions (excluding outliers and extreme cases in terms of site conditions).

The design reference level D.R.L's were calculated using candu sites specific data in Canada outlining actual tritium routine releases and field validation on tritium concentration. Exposure pathway conditions including dairy, forage, and livestock farms, as well as highly diversified animal husbandry farms (e.g., with pigs, poultry, etcetera) and the presence of rural populations on drinking water wells and in small villages, all in proximity of candu N.P.P sites. The presence of recreational water bodies near the site and, consequently, fishing activities, were also factored in.

This modeling is only for routine, planned, and controlled releases and covers long-term and long-range (transport distance) consequences for normal (planned exposure situations). As mentioned previously, modeling for the maximum unplanned and uncontrolled releases of tritium due to significant load events from external (e.g., earthquakes) and internal (e.g., tritium fire, explosion, human error) hazards is totally different and done with other modeling methodologies and relate to “emergency exposure situations” for which public dose thresholds are different and temporary. It is worthy to note that Type One's limit for its “Level 0 Design” basis is 3 kilograms of H 3 , which is a bounding limit to various siting options (excluding outliers and marginal cases) to obtain regulatory exemptions for the development of off-site emergency planning and responses given that such release limit is bounding the dose projection below the 1 rem (10 millisievert) dose projection threshold. This was calculated assuming a 100% H.T-H.T.O conversion.

Table 3 presents the exposure pathways and physical characteristics of the design reference site for the calculated tritium D.R.L's limit of approximately 30 g (289,500 Curies) and 300 g (2895,000 Curies) of tritium for meeting regulatory requirements in terms of dose constraint for air emission and general public dose limit and of tritium (respectively 0.1 millisievert and 1 millisievert). The last column provides some considerations and an understanding of the characteristics specific to the reference site (hypothetical site based on conservative assumptions) and identifies the pathways that could be used, when identified as exclusion criteria (potential disqualifying factors) for the selection of potential sites. Table 4 presents a reasonable set of soils, plants and animals as a benchmark for this model, given that levels of diversity vary greatly depending on geography and environmental conditions. These parameters will therefore be expanded, as necessary, to reflect local conditions.

: Tritium exposure pathways and the key variables affecting dose for the T1E SFPP design reference levels. The foundational pathway is inhalation dose, which is highly correlated to the effective height of the release point, set at a minimum of 60 meters, as well as dominant winds and atmospheric conditions; this pathway influences all others. Other significant pathways include water immersion and ingestion from drinking water wells, where the absence of wells within 0.5 to 3 kilometers downwind can substantially decrease the total dose. Animal and dairy product ingestion pathways are categorized by the transfer of HTO or Organically Bound Tritium through ponds, soil pore water, forage, or direct air-to-animal transfer. Specifically, pathways involving milk and dairy products are identified as particularly important for infants and children. Finally, the ingestion of local fruits and vegetables via air-to-plant transfer is a key consideration, with exposure potentially reduced by the absence of local farms, though local gardening and wild fruit harvesting remain relevant factors.

Table 4 summary: The specific input types considered in the model, categorized by source. These include four soil types: Sand, Loam, Clay, and Organic; and two sediment types: Fresh water (lake, pond, river) and Marine. Terrestrial animals are represented by ten inputs, including Milk, Beef (including veal, liver and offal), Pork, Lamb, Poultry, Egg, Deer, Rabbit, Wild waterfowl, and Honey. Aquatic animals include Freshwater fish, Marine fish, Freshwater invertebrates, and Shellfish. Terrestrial plants cover Forage (for animals), Grain (for human and animals), Fruits and vegetables, and Potatoes (representing root crops), while aquatic plants consist of Dulse (seaweed) and Freshwater plants.

3.3.3. Site-Specific Calculation of D.R.L for T.V.A's Bull Run Site in Claxton, T.N

At the time of publishing this paper the analytical efforts were still underway to finalize the Derived Release Limits (D.R.L) for T.1.E S.F.P.P Infinity Two routine tritium release for Reference Person(s) located near the Bull Run Facility (B.R.F) on public accessible land, using realistic exposure pathways. Preliminary results discussed with regulatory authorities suggest limits well above as those established for the design reference level of 30 g: 60 to 70 g of tritium in terms of maximum annual atmospheric releases for meeting the public dose constraint of 0.1 millisievert and 600 to 700 g for meeting the public dose limit of 1 millisievert.

Sensitivity analysis on applicable exposure pathway parameters was still in progress. This is entirely logical given the absence of many assumed exposure pathways in the conservative hypothetical scenario and the dominance of inhalation (exposure pathway No 1 in Table 3) in terms of contribution to the dose and from a few other exposure pathways only (consumption of wild fruits and locally grown animals such as poultry).

Most of the remaining work involves sensitivity analysis of the parameters applicable to exposure pathways retained in the modeling to refine the values assigned to each variable so that they correspond as closely as possible to reality in the field. At this stage of the analysis, no major changes are expected, only calibration and minor variations in the results. Nothing that could influence the design of tritium fuel cycle systems, given that D.R.L's are a regulatory value (release acceptance criteria) and not a discharge target for the S.F.P.P operations as the design objective for optimized fuel economy and public safety is no routine release.

3.4. Identifying Radiation Safety Constraints and Requirements for Relevant Infinity Two S.S.C's

T.1.E design activities are aligned with radiation safety objectives and principles with a goal of obtaining an operating license involving S.S.C's relevant to radiation safety to meet dose limit requirements for the public and workers, with alara as an overarching objective. The alara principle is an integral part of the regulations in the U.S.A and § 20.1101 (b) Licensee shall use to the extent practical procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable.

Regulatory expectations are very clear regarding S.S.C's relevant to radiation safety to keep doses received by workers and the public within regulatory limits and constraints values. Although no quantitative values are prescribed for their performance and reliability, as is the case for nuclear power plants, S.S.C's will be analyzed in all modes of operation, including maintenance, decommissioning and failure modes (e.g., using the master logic diagram and the functional failure modes effect analysis from the early design stages) to demonstrate that they meet safety objectives before construction begins. This analytical exercise is not only for defining radiation safety requirements, but also to ensure that the design requirements also cover assets protection and physics constraints. The outcome of this analysis informing the design will help the Tennessee regulatory body to build its inspection list to be used for the final inspection prior to commissioning and start-up of the S.F.P.P.

T.1.E's S.F.P.P (F.O.A.K and N.O.A.K's) design activities take these possible failures into account in addition to the loads resulting from normal operating conditions that can have an impact on the S.S.C's relevant to radiation safety. The design process systematically includes regulatory constraints to meet key associated safety objectives for each relevant S.S.C's.

3.4.1. The master logic diagram (M.L.D) and functional failure mode effect analysis (F.F.M.E.A) joint approaches for identifying radiation safety constraints in early design phases

The Master Logic Diagram (M.L.D) is a safety analysis method that can be used for identifying the hazards (loads) that may affect the performance and reliability of fusion machine systems relevant to radiation safety. While the bottom-up Functional Failure Mode Effect Analysis (F.F.M.E.A) method depends on the availability of more detailed functional engineering system diagrams, the top-down M.L.D approach can work earlier in the design process using less detailed functional engineering system diagrams (for example, with limited system definitions and basic functional block diagrams). Both the F.F.M.E.A and M.L.D are useful methods for early design activities to ensure radiation safety considerations are taken into consideration and built-in rather than added-on later in the design process.

The M.L.D is useful for identifying the interfaces (movement of information or materials) between the functional systems.

The main stages of the method can be summarized as follows:

1. Identification of the main events that could lead to the partial or total loss of control and containment of radiation or radioactive material at the level of a functional system (e.g., loss of integrity of the first radiological containment barrier represented by the vacuum vessel as expressed in figure 5 which presents a well-known generic example applicable to many magnetic confinement fusion technologies).

Figure 5 summary: A side-by-side diagram comparing the conceptual frameworks of Master Logic Diagrams (MLD) and Failure Effects and Criticality Analysis (FFMEA). The MLD flow on the left progresses from sub-events, such as cooling pipe loop breaks or leaks, up to a postulated main event involving a breach of the vacuum vessel. The FFMEA flow on the right organizes a similar sequence from process functions and failure modes up to a postulated basic event resulting in the transfer of materials to the expansion volume. The diagram illustrates how these two methodologies map different logical structures to the same underlying design hazards.

2. The main event is then broken down into sub-events that may contribute to or explain the occurrence of the main event (main event

e.g. confinement failure of the vacuum vessel). These subevents are broken down by linking key interfaces between functional systems and components (e.g., hydrogen or non-hydrogen species leak rate at various vacuum vessel interfaces with other systems: in-vessel cooling systems, ex-vessel cryostat, etcetera). In this analysis, the possible physical parameters (loads and consequential impacts) are more important than the deterministic analysis of failures and transients. As the failure modes are presumed using engineering judgement of the systems and their loads. The M.L.D approach allows logic rather than deterministic analysis to inform the subsequent advanced design requirements and the S.S.C's; deterministic analysis is then utilized later in the design process when more practical.

3. Once this exercise has been completed, a list can be made of the initiating or basic events that cannot be divided into sub-events. Also, the radiological safety constraints and requirements (detection and control systems for leak rate and dose rate) can be identified and inform the more advanced design stages of each S.S.C's relevant to radiation safety.

The F.F.M.E.A analysis can then be completed in the reverse order to complete the list of these initiating events on the same basis as for the M.L.D approach. that is, without all the design details available to allow more specific evaluations at component and sub-system levels necessary to investigate failures and/or challenging plant transients that could affect system functionality.

As an example, starting with a process function (e.g., preserve integrity and leak tightness of the vacuum vessel), the corresponding P.B.S element (vacuum vessel) is selected for F.F.M.E.A and a loss of function is postulated to trigger a failure (e.g. loss of coolant from the in- vessel cooling system inside the vacuum vessel) followed by the consequential impact of pressure rising in the vacuum vessel thus compromising the structural integrity of the radiological containment barrier. A crack or rupture in the first radiological containment barrier would lead to the second barrier, the ex-vessel having an abnormal dose rate reading. This would indicate a potential breach to the vacuum vessel. Design features need to be built in to prevent the vacuum vessel's pressure from fluctuating outside the safety margins (monitoring of vacuum integrity/pressure changes and/or pressure relief systems).

The M.L.D and F.F.M.E.A, carried out in parallel or in sequence enable a complete list of initiating events for the early design phases (pre-conceptual, conceptual and preliminary design phases). This analysis informs the design by determining the requirements of the process (DiD level 1) plus informing the detection and control systems (DiD level 2) in an optimal way.

The leading assumption in this exercise, which becomes a design rule, is that the functional requirements of all relevant engineering S.S.C's are meeting asset and investment protection goals along with radiation safety (regulatory) constraints. Based on this no other supporting systems or design features - active or passive - are necessary for any operational modes (start-up, steady state operations, shut down, cyclical and preventive maintenance activities, etcetera).

By adopting this philosophy, Type One Energy can assert that safety is built into the design, that no additional engineering features will be required and/or demanded during the final stage of the licensing process, and that this holistic approach will go shape a modern and effective safety assessment process supporting licensing.

Beyond the control and containment of mobilizable radiological inventories and the use of the M.L.D followed by the F.F.M.E.A method, regulatory constraints will also be identified to optimize the design of Infinity Two. These constraints are related to alara principles and the reduction of in-vessel and ex-vessel activation levels using radiological shielding and a targeted quality program for material selection.

Early use of the Monte Carlo Radiation Transport Code OpenMC for material activation analysis offered a practical, open-source alternative to traditional codes like M.C.N.P by combining neutron transport and activation calculations in a single environment. Unlike M.C.N.P, which typically requires external tools for activation studies, OpenMC performs depletion during irradiation and tracks the evolution of nuclei inventories over time. After depletion, it evaluates the specific activity of each radionuclide and compares these values against specific activity limits (S.A.L's) for near surface disposal defined by Fetter (1990) and N.R.C thus informing waste classifications to optimize material options and identify impurities limit as to not exceed G.T.C.C thresholds. This capability enables direct waste classification from simulated irradiation history without relying on separate post-processing steps.

Initial analyses for T.1.E Stellarator design and associated material selection indicate that the production of intermediate-level waste (I.L.W), or its closest U.S. waste classification, greater than class C (G.T.C.C), is not expected even when accounting for well-known material impurities in structural materials such as steel and neutron multiplier materials. Nevertheless, material impurities remain a dominant factor in determining waste classification, as well as the resulting decay heat and radiological activity. To address this sensitivity, ongoing work is focused on optimizing material selections and actively engaging with vendors to better control and, where possible, reduce impurity levels to meet project activation and waste management requirements.

In addition to material selection and optimization, OpenMC's ability to compute activation and classify waste from depletion results supports early engagement with recycling partners. By modeling decay profiles for irradiated components, engineers can identify alloys suitable for reuse and plan processing strategies that comply with regulatory requirements. This approach reduces lifecycle costs and aligns with sustainability goals while maintaining radiological safety standards. Recycling approaches will be employed where technologically feasible and economically viable, and the remainder of material will be kept within L.L.W disposal limits with appropriate material choices to stress the environmental value of fusion in utilizing natural assets efficiently, assert the fundamental premise of fusion as a nuclear energy source with minimal environmental impact, and gain public acceptability for fusion.

Finally, OpenMC integrates transport, activation, and dose-rate validation in one workflow, streamlining shielding design, and zoning for controlled, restricted, and unrestricted areas. These simulations feed directly into defining radiation safety constraints for S.S.C's embedding regulatory and operational requirements early in the design process rather than late-stage corrections. Compared to M.C.N.P workflows, which often depend on multiple external codes, OpenMC provides a unified and transparent platform that accelerates design decisions and ensures compliance.

When performed from the earliest phase of design and integrated with other non-radiological constraints (such as mitigating industrial hazards and asset protection, through a Model-Based System Engineering (M.B.S.E) platform) these analyses pave the way for a better integration of all the constraints into comprehensive S.S.C's requirements for informing the final design.

4. Implementation of the holistic safety-by-design concept

4.1. Stellarators and inherent safety: a physics-driven advantage

The inherent macroscopic stability of the S.F.P.P plasma core relative to the more common tokamak-based core provides important physics-driven advantages in the pursuit of radiation safety-by-design. In particular, the S.F.P.P plasma core greatly reduces the frequency and magnitude of tran-zee-unt thermal and mechanical loads to the vacuum-vessel and in-vessel components. The advantages include a drastically reduced risk of damage to in-vessel components and cooling lines and reduced dust production.

Disruptions represent major operational and safety challenges for tokamak-based fusion machines. A disruption is a sudden, uncontrolled loss of plasma confinement caused by the growth of large-scale instability. During a disruption, the plasma temperature collapses on the scale of a few milliseconds (“thermal quench” and the plasma current rapidly decays within tens to hundreds of milliseconds (“current quench”), depositing large thermal loads and inducing strong electromagnetic forces on the vacuum vessel and internal components.

These events can melt plasma-facing armor, deform structural elements, and generate runaway electrons capable of causing localized, penetrating damage down to embedded cooling lines. Disruptions are primarily a tokamak-specific problem for three reasons. First, high-performance tokamaks are inherently vertically unstable, so that the position of the plasma within the vacuum vessel must be constantly actively stabilized by external control coils.

Failure of vertical stabilization, either due to hardware failure or internal plasma evolution causing the plasma to exceed controllability limits, results in the plasma drifting rapidly to the first wall and disrupting. Stellarators, by design, do not suffer from such a pervasive large-scale instability. Second, tokamaks rely upon a large toroidal electric current driven in the plasma (of order 10 M.A in a fusion power system) for confinement and stability.

That current can drive a wide variety of current-driven instabilities that can cause disruption, and when that current collapses, severe mechanical and thermal stresses are unavoidable. Stellarators, by contrast, generate their confining magnetic fields almost entirely with external coils and therefore do not experience the same catastrophic current-driven failure modes. Finally, the pressure-driven instabilities common to both designs, while readily avoided in each, exhibit a softer evolution in stellarators that almost never leads to disruption.

High-confinement mode (H-mode), while desirable for achieving reactor-relevant performance in tokamaks, introduces additional risks through edge-localized modes (E.L.M's). H-mode forms a strong edge transport barrier that improves energy confinement but drives steep pressure gradients at the plasma edge, which periodically relax through E.L.M's-bursts that eject particles and heat along magnetic field lines to the divertor, producing intense tran-zee-unt heat loads that can significantly limit component lifetime in machines such as iter. Large E.L.M's are a critical concern for tokamaks operating in steady H-mode, necessitating complex mitigation or suppression schemes. In contrast, steady state I.N.F.2 stellaror operation will not involve H-modes and E.L.M's.

Dust in fusion machines is a critical issue, as it can introduce impurities that degrade plasma performance and pose safety concerns related to chemical reactivity, tritium retention, and radioactivity—which can complicate the licensing process—as well as its potential to interfere with in-vessel plasma-facing diagnostics.

In tokamak systems, edge-localized modes (E.L.M's) generate tran-zee-unt, high-intensity heat and particle fluxes to plasma-facing components (P.F.C's), intensifying erosion and dust release through mechanisms such as thermally induced brittle fracture of surface materials, detachment (flaking) of redeposited layers, and localized unipolar arcing driven by strong tran-zee-unt electric fields. These processes can degrade plasma purity and accelerate P.F.C wear, thereby influencing component lifetime and safety margins. While E.L.M-free operating scenarios are available to tokamak systems (see for example I-mode, Q.H-mode, L-mode, and negative triangularity, in practice most power system designs presently baseline to the elming H-mode due to its robust confinement characteristics, that is EU-demo, A.R.C, and C.F.E.T.R.

In contrast, the stellarator configuration employed in the I.N.F.2 S.F.P.P is immune to dust-production pathways associated with plasma disruptions and E.L.M-driven tran-zee-unt loading. Accordingly, such mechanisms are not expected to contribute appreciably to dust generation in the I.N.F.2 vacuum chamber.

In addition to reducing dust generation, these characteristics substantially reduce cyclic loading and fatigue-related stresses, thereby enhancing structural reliability and overall system safety.

4.2. Engineering constraints that bound radiological risk

A fusion machine designed under the safety-by-design approach will have radiological design constraints integrated with the non-radiological constraints to inform the design of S.S.C's. This suggests that focusing on functional and structural integrity for the control and containment of radiation and radioactive materials will, by design, guarantee integrated radiological safety in all relevant S.S.C's.

The radiation safety constraints identified at the functional conceptual design phase level lead to early considerations on the need for specific detection (instrumentation) and control systems to meet regulatory dose constraints and limits. For example, superconducting magnets producing large Lorentz forces and a quench event releasing significant heat energy can both potentially cause structural damage. Unlike fission equipment, structural integrity is largely independent from pressure loads; however, accidental pressure loads on safety-significant components and rooms will still be evaluated within the applicable load combinations to ensure compliance with design requirements.

Varying heat fluxes can have a significant impact on the divertor and first wall. Uncontrolled heat fluxes from the plasma may cause melting or cracking of the divertor tungsten plasma facing surfaces. In T.1.E's S.F.P.P, the control system will utilize feedback-controlled injection of neutral gas to regulate the heat flux to the divertor.

The injected neon radiates energy and undergoes charge-exchange reactions at the plasma edge, reducing the localized heat and plasma particle flux to the divertor. This "detachment" process protects the divertor plasma facing surfaces from damage even under extreme thermal loads.

Thermal loading and irradiation-induced material aging are anticipated operational phenomena that are addressed through conservative design limits and remote replacement strategies defined for component lifetimes in-vessel systems. These effects are managed within the plant safety framework through qualification analyses, operating envelopes, and surveillance assumptions applied during design.

In-vessel components are configured to support a sector-based maintenance strategy, allowing scheduled inspection and replacement activities to be executed in a controlled and localized manner. Maintenance planning is fully integrated into the plant availability model, ensuring that degradation mechanisms do not challenge structural integrity or safe operation. (See Section 5 on maintenance)

Through this integrated approach, component aging and replacement are treated as planned lifecycle activities, providing confidence that safety functions and operational availability are maintained throughout the facility's lifetime.

Once radiation safety, engineering and physics constraints are identified through the M.L.D and F.F.M.E.A/F.M.E.A processes, integrated S.S.C's design requirements are identified for all S.S.C's. For example, from the first wall (which is located at the innermost layer) to the outermost layer of cryostat, every structural and functional component plays a crucial role in maintaining the integrity and performance of the fusion system. The potential failure modes of each S.S.C are systematically analyzed to assess the consequences of these failure modes through an event-tree structure and to identify appropriate design requirements that control the radiological consequences of any failures and contain the mobilizable radiological inventories within predefined radiological physical boundaries or barriers.

Particularly significant are failures resulting from material degradation (from irradiation), which we expect to occur at predictable locations based on our understanding of mechanics and service conditions. These anticipated failure zones are intentionally designed to localize damage and prevent it from spreading beyond the defined containment area. Similar approaches will also be incorporated into time-limited aging analyses, which account for the cumulative impacts of thermal and mechanical cycling, neutron exposure, and material embrittlement. These calculations must be evaluated to cover the extended operating period to inform the aging management strategy.

One of the primary objectives of T.1.E preventive maintenance program is to conduct maintenance activities before failures are expected to occur. By proactively addressing maintenance needs, the risk of unexpected downtime can be minimized, ensuring the continued reliability of operations.

Based on current estimates, a broad range of materials used in operations are expected to have an optimal service life between 4.5 and 5 years under anticipated radiological conditions. Understanding this timeframe allows for more effective planning and scheduling of shutdowns aligned with maintenance requirements.

However, data regarding the degradation of material performance and reliability due to radiologically induced aging is currently incomplete. To address this uncertainty, reference samples will be collected periodically. These samples will be analyzed to track the progression of failure over time, providing valuable insights into material aging and deterioration.

The information gathered from monitoring reference samples will be instrumental in refining maintenance and shutdown plans. As more data becomes available, strategies will be adjusted to ensure maintenance remains proactive and effective, further reducing the likelihood of failures and optimizing operational efficiency.

Preventive maintenance for fusion systems must be integrated from the design phase to ensure that all components remain serviceable throughout their operational life. Maintainability is ensured through modular design, standardized joining methods, and alignment features compatible with remote-handling systems, enabling routine preventive tasks to be executed without direct human access.

During operation, preventive maintenance relies on periodic inspection using in-vessel viewing systems and remote tooling to assess condition, thermal performance, and wear of plasma-facing and structural components. Condition monitoring through temperature, strain, and coolant-flow rates supports early detection of degradation. Components are accepted for installation only after demonstrating functionality through validated procedures, ensuring that every installed module can be operated and replaced within defined limits. Together, these measures form a practical preventive-maintenance framework suitable for I.N.F.2.

This strategy is a key element of the broader level 2 defense-in-depth and investment protection framework, ensuring that operational failures do not compromise radiological safety or the functionality of adjacent systems.

4.3. Role of fusion codes and standards supporting safety-by-design approach

In industry, a large part of radiation safety and asset protection assurance relies on rigorous engineering processes and structured quality systems—both supported by well-established codes, standards, and their underlying objectives. For the S.F.P.P, the iterative development and progressive adoption of fusion-relevant Codes and Standards will be based on the technical objectives of Infinity Two. These technical objectives, together with the practical attributes of a quality assurance program, will be derived from several decades of R&D experience that contributed to the maturation of fusion S.S.C's and from other integrated requirements specific to a commercial fusion power plant.

It is important to note that, in the United States, the Nuclear Regulatory Commission's regulations, specifically 10 C.F.R Part 30 and the associated guidance in NUREG-1556, Volume 22 do not prescribe or impose any specific codes or standards. This is because the licensing framework applicable to fusion facilities does not involve design certification and therefore does not mandate compliance with a predetermined set of technical codes. As a result, the adoption of codes and standards in the fusion sector is pursued primarily for investment protection, engineering reliability, and the safety of workers and plant personnel, rather than as a direct regulatory requirement.

Housed within the Stellarator core are individual systems such as Fusion Superconducting Magnets, Vacuum Vessel, Blanket, First Wall, and Divertor, each with its own complexities—not to mention the intricate design and safety considerations that arise when they are assembled and operated together. Inside the vacuum vessel, plasma reaches the highest temperatures in the solar system, about 150 million degrees Celsius (compared to the sun's core temperature of roughly 15 million degrees Celsius). Just a few meters away from the plasma, superconducting magnets need to be cooled to near the temperature of deep space. Beyond temperature, numerous physical phenomena occur in the Fusion Core components, making it more complex in some respects from an engineering standpoint than a fission reactor.

There are no other engineering systems from which fusion engineers can borrow code for all design aspects, so core fusion components require their own dedicated code development. Although the physics and operation of a fusion machine are in some ways more complex than those of a fission reactor, the fusion process also guarantees inherent radiation safety, unlike fission, which carries a risk of runaway reactions and meltdown. Due to these fundamental differences, fusion code developers should not adopt any principles related to fission but must approach the code with an unbiased perspective to ensure a safe and unaffected developmental environment.

A comprehensive code framework development must advance in parallel with the technological maturity of fusion systems; otherwise, overly prescriptive requirements may constrain progress. This principle is exemplified by the French nuclear code R.C.C-M.R.x, which introduced probational rules for newly incorporated fusion materials in recognition of the limited industrial feedback available at the time. In the same manner, a commercial fusion code comprehensive framework should incorporate mechanisms that allow for flexibility and staged adaptation until the fusion industry establishes a sufficient base of long-term operational experience.

On the other hand, outside the core fusion systems (such as the in the T.1.E's S.F.P.P Core), the Balance of Plant, and auxiliary systems can potentially adopt existing Codes.

Taking the example of irradiation-compatible steels, such as Reduced Activation Ferritic Martensitic (rafm) steels, R.C.C-M.R.x has been working on code qualification of Eurofer97 for the past few years. Once fully established, the Code will require the associated industries in the supply chain to comply with the Code specification for the material, from production, testing, to the product marking requirements during delivery. As a similar effort, the U.S Department of Energy (D.O.E) has funded the A.S.M.E code qualification of Castable Nanostructured Alloys (C.N.A's) in early 2025. Although these are excellent catalysts for fusion industrialization, other implementation aspects of the fusion code-informed comprehensive framework activities, such as welding and structural integrity testing, currently have lower and varying levels of maturity. They will need to rely on existing codes and standards until a comprehensive fusion-specific code that defines a fusion supply chain is developed.

The reliability of a robust design process based on radiation safety constraints integrated with engineering and physics constraints (thanks to a supply chain that meets integrated requirements) can be illustrated by an example.

At one end of the development pipeline, design engineers articulate the structural and functional requirements that define the intended performance of the system. At the other end, realizing these requirements demands a systematic set of processes supported by a qualified supply chain capable of adhering—at least in the early stages—to relevant portions of existing codes and standards. The design process is therefore intrinsically linked to the supply chain through multiple interconnected layers of material procurement, fabrication, inspection, and verification, each of which must consistently demonstrate conformity to documented requirements.

For example, considering the failure modes discussed earlier, during the fusion engineering process, design engineers depend on the specified material quality and the mechanical and physical properties of materials from the early design phases. It is the responsibility of a well-managed Q.A/Q.C program and a reliable supply chain to meet these expectations, so companies can build a fusion island as designed and function as intended. In this context, codes like A.S.M.E and A.W.S act as a thread connecting material manufacturers, part fabricators, welding and joining companies, non-destructive examination experts, inspectors, and other key supply chain contractors, leading to the integration of a fully functioning machine in a well-coordinated, repeatable, and traceable manner. Implementation of this process, which has evolved through decades of correction and refinement efforts, will minimize material and fabrication flaws from impacting the fusion machine reducing the possibility of failure. While relying on the existing codes and supply chain for current needs, it is important to support the development of a fusion-relevant code and supply chain, as iter demonstrated based on its code books like R.C.C-M.R.x. (commonly referred to as iter-qualified vendors') However, unlike the iter supply chain, a fusion supply chain that supports commercial fusion deployment will not be driven by regulatory stringency (in the U.S.A), but by the need to accommodate varying engineering complexities and to adapt rapidly to the fast-evolving nature of fusion technology—absorbing lessons from first-generation machines and quickly applying them to the second generation as the technology advances at unprecedented speed.

4.3.1. Use of existing codes and standards for fusion engineering

Key documents such as R.C.C-M.R, R.C.C-M.R.x, and iter S.D.C-I.C are valuable, fusion-oriented codes and structural standards created for iter and now used in the design of the European demo fusion system. Although these frameworks provide extensive technical guidance to the S.F.P.P, they do not fully address the needs of commercial fusion systems, which require not only efficiency, scalability, and flexibility but also an economically sustainable approach to engineering integrity to support reliable long-term operation. At the same time, A.S.M.E Section 3 Division 4 published its first edition in 2023 and is aiming to become the most streamlined and dedicated codebook for commercial fusion machines, incorporating input from public institutions and private fusion companies. However, Division 4 remains in its early stages and will need significant development, with its progress closely linked to the evolving technical requirements and feedback from industry stakeholders.

Although Division 4 is written specifically for fusion applications, it currently resides within the broader fission-oriented framework of A.S.M.E Section 3. This placement is misaligned with the reality that fusion and fission power systems share virtually no common components and rely on fundamentally different scientific principles, engineering design philosophies, and safety approaches. Consequently, establishing a dedicated Fusion Section—such as a proposed Section 14—would enable engineers to approach fusion technology with a clean, purpose-built framework, free from the legacy assumptions embedded in fission-based code structures.

Such an independent A.S.M.E section would be particularly advantageous during today's period of accelerated fusion technology development. A minimalistic, fit-for-purpose, and adaptable code—focused on essential requirements rather than inherited constraints—will better support commercial fusion enterprises such as Type One Energy, enabling them to innovate rapidly while maintaining appropriate safety and quality standards.

While existing pressure-vessel and structural codes—such as A.S.M.E Section 8 Divisions 1 and 2, A.S.M.E B31.3 for process piping, and A.W.S D1.1 and D1.6 for structural welding—can be provisionally adapted for interim use with applicable test benches for validating suitability for service, it is important to support the development of fusion-specific codes for the long term. In addition to contributing to the Division 4 fusion Code development committee, Type One Energy is also an industrial advisor for the D.O.E-funded FIRE-IMPACT project, which focuses on the A.S.M.E Code qualification of Castable Nanostructured Alloys (C.N.A's), a potential candidate for the First Wall and Blanket components ^{} .

4.3.2. How to develop fusion-specific codes and standards for safety

The safety considerations arise from an accurate understanding of relevant failure mechanisms. An industrial Fusion Code book must reflect the actual failure modes and degradation mechanisms unique to fusion. Fusion environments involve 14.1 megaelectronvolts neutrons, which are far more energetic than fission neutrons, resulting in unique material degradation mechanisms such as transmutation-induced embrittlement and helium bubble formation. Radiation-induced swelling, material property degradation, embrittlement, and related damage mechanisms like fast fracture and ratcheting are some of the material and design challenges. Understanding how materials behave under irradiation and high temperatures over a machine's operational period enables engineers to address safety at multiple levels.

Fig. 6 below shows the major elements a fusion Code book needs to cover to allow for design with safety in the engineering process.

Figure 6 summary: A diagram showing a central core of codes and standards ensuring safety by design, with arrows radiating outward to six key operational areas: QA/QC, Inspection, Manufacturing, Documentation, Material, and Design. The diagram illustrates that these codes and standards are integrated across all primary stages of the production and quality control lifecycle to ensure overall safety.

Typical code structures are built around the fundamental elements of materials, design, and welding, all of which ensure structural integrity. These are supported by quality assurance and quality control (Q.A/Q.C), inspection requirements, manufacturing practices, and associated documentation.

4.3.3. Fission codes cannot provide fusion safety

It is widely recognized that the safety profile of fusion machines differs significantly from that of fission reactors. This fundamental difference renders the fission codes such as A.S.M.E Section 3 Divisions 1, 2, 3, and 5, along with N.Q.A-1, [55] unsuitable for fusion applications. Despite their proven track record in fission systems, these codes were intended to address the unique failure modes, operational conditions, and safety considerations inherent to fission. The core philosophy of the fission codes emphasizes high-integrity pressure boundaries and components to provide assurance that the reactor can maintain the configuration of the fuel and its control rods, in addition to moderating the temperature of the reactor.

Attempting to repurpose parts of fission-oriented codes and standards for fusion applications results in a misapplication of resources and incomplete coverage of fusion-specific needs. Applying fission-centric engineering approaches and safety factors leads to overly conservative designs that are cost-prohibitive without improving safety. This highlights the necessity of developing a dedicated quality system tailored to fusion. Such a system will need to evolve alongside the industry, but it must be rooted in a robust and widely understood baseline to facilitate engagement with the existing supply chain.

One potential candidate for developing this foundation is the quality system defined in A.S.M.E Section 8 Division 1 ^{} , which governs unfired pressure vessel construction. It offers a practical, scalable, and accessible starting point, from which a fusion-specific quality framework can be gradually built and refined through operating experience.

It is also important that fusion-specific code development must take place in an environment that is respectful and honest about the evolving maturity stages of fusion technologies with the involvement of private and public fusion companies. A typical Industrial Code development can take 5 to 10 years to reach a stable form, and then it keeps evolving with the newer technologies and needs learned from the actual operating experience from the users of the code rules. To manage immediate needs, U.S Codes such as A.S.M.E may follow the R.C.C-M.R.x pathway and include probationary and fast-track pathways for materials under development.

5. Special considerations for radiation safety constraints and requirements for maintenance activities

One of the key challenges arising from the lack of operating experience under fusion-level irradiation is predicting the maintenance window and end-of-life periods of S.S.C's. While modeling and extrapolations from data on fission-irradiated materials, along with decades of research in irradiated materials, help predict the nature and location of failures, that data does not include significant history with 14.1 MeV neutrons. Based on this, online monitoring of component degradation in the fusion machine will provide valuable feedback to update those predictions and refine the maintenance intervals for replacement of different S.S.C's.

5.1. The prominent role of maintenance to support licensing for radiation safety by design and operating programs

Cyclical or preventive maintenance activities are key contributors to provisions for radiation safety and are particularly prominent in pre-licensing and licensing discussions and engagement efforts with regulatory authorities.

From one angle, and as we have seen in all previous sections, several essential elements contribute to the “safety-by-design” approach. This spans from radiation safety constraints informing design requirements for all pertinent S.S.C's, to the material selection process and activities related to procurement, manufacturing, and fabrication of assemblies.

Preventive replacement or repair of radiation safety-relevant components and systems is imperative in mitigating issues associated with aging and potential performance decline. These components and systems may experience diminished effectiveness over time, thereby negatively impacting the quality of the plasma and machine performance by extension.

Routine preventive maintenance is critical to sustaining optimal system performance. By systematically replacing or repairing parts before degradation escalates, organizations can minimize erosion from and thereby minimize the source of loose contamination. This proactive strategy is vital for upholding safety standards and reducing the probability of radiological hazards. Continuous monitoring, inspection, and maintenance are critical to ensure that the negative impacts of aging do not compromise the safe operation of the stellarator or its ability to control and contain radiation.

During these maintenance activities, comprehensive measures are implemented to ensure the protection of personnel from radiation exposure. This includes proven measures that have been utilized in commercial nuclear power for decades such as radiation monitoring, radiation work procedures, personnel protective equipment (P.P.E), and radiation worker training.

These cyclical maintenance activities will also be of major importance for gathering valuable information to feed operational experience databases for the industry and regulators. This operational experience on the behavior of aging and reformulated S.S.C's under operating conditions feeds into safety assessment validation, the development of fusion-specific codes and standards and the potential reformulation of regulatory requirements that are better targeted to the operational reality of fusion machines that sets the foundation of a mature and sustainable fusion industry.

From another angle, maintenance operations and their often-intrusive nature into highly restricted areas in terms of radiation safety, are those that require special and adapted standard operating procedures (S.O.P's) to ensure compliance with regulatory dose limits and constraints for workers and the public while adhering to the alara principles. Maintenance activities may involve the mobilizing and interim storage of S.S.C's for repair or refurbishment, including exposure to and replacement of irradiated components. Shielding approaches during these operations will be detailed as part of the maintenance program adhering to radiation safety constraints, requirements for licensing and asset protection. This may include wet and/or dry shielded storage areas to cool and/or isolate recently removed radioactive materials.

This program must also address threats of contamination and radiation to the plant, workers and the public during handling and refurbishment. Replaced activated material can be loaded into specially designed liners or casks for transport to radioactive material processing facilities for recycling, reclassification, reduction and/or disposal.

The N.R.C licensing application guidelines for fusion machines devotes an entire section to this topic covering requirements and considerations to be included in license applications that go beyond those prescribed for particle accelerators. It requires commitment and evidence for the provision of adequate resources (including space, equipment, time, personnel and contractors) to the radiation protection program to ensure the public and workers are protected from radiation hazards and compliance with regulations is maintained during routine and cyclical maintenance activities.

5.2. Radiation safety hazards assessment informing early development of cyclical maintenance program

The following sources of radiological risk are applicable, requiring the integration of various protective measures into the design and operational radiation protection program to support maintenance activities:

1. Plasma sources of prompt neutrons and x-rays and gamma rays from neutron activation;

2. Gamma dose rate from neutron activated S.S.C materials;

3. Potential inhalation or ingestion of radionuclides from activated dust, and inhalation/skin absorption from parasitic losses of tritium.

Maintenance activities conducted during scheduled shutdowns in radiation areas necessitate a rigorous approach to protecting worker health and ensuring compliance with prescribed occupational dose constraints, limits and other regulatory requirements in terms or radiation safety programs. To meet these critical objectives, comprehensive and dedicated radiation safety Standard Operating Procedures (S.O.P's), detailed Radiation Work Plans (R.W.P's), and specialized training programs must be systematically developed and put into practice.

The implementation of these radiation safety protocols draws upon established methodologies and best practices refined in commercial nuclear power generation, naval nuclear operations, and significant research and development projects, including those undertaken at J.E.T's, iter, and mast. By leveraging the collective operational experience and procedural knowledge from these sectors, maintenance processes can be optimized to uphold the highest standards of radiation protection for all personnel involved.

Upon shutdown, the radiological hazard originating from the plasma source is eliminated, as plasma operations cease and no further prompt radiation is generated. Simultaneously, the secondary hazard—primarily photons emitted from materials within the machine that have become activated during operation—begins to diminish. This decrease in radiological risk is attributed to the natural process of radioactive decay, which reduces the intensity of photon emissions over time.

To ensure the safety of maintenance personnel, all maintenance activities are scheduled to commence only after an adequate cooling period has elapsed following shutdown. During this interval, remote interaction with the machine is prioritized, with the stellarator remaining closed in order to limit personnel exposure. Access by workers to areas near the stellarator is intentionally restricted for several days post-shutdown. This precaution is supported by research indicating that the radiological conditions inside the machine improve significantly—by one to two orders of magnitude—within just seven days after shutdown.

After shutdown, small quantities of tritium (referred to as hazard 3) may remain in the gaseous form—either as H.T or T2—within both the Stellarator and the Stellarator Hall. This residual tritium is primarily a result of permeation and outgassing from the internal surfaces of the vacuum vessel as well as components involved in the tritium fuel cycle's S.S.C's.

To address this, the facility employs a dedicated gas collection system within the hall. This system is designed to efficiently collect, control, recycle, and, when necessary, safely release tritium gas. Functioning continuously throughout both operational and shutdown periods, the gas collection system ensures that the concentration of gaseous tritium remains minimal, particularly during maintenance activities, thereby contributing to the overall safety and radiological protection of personnel and the environment.

The S.F.P.P will be assembled from multiple pie-shaped segments, or modules, each engineered with the capability for straightforward replacement. During maintenance activities, the removal of any individual segment results in direct exposure to a highly irradiated internal environment.

To ensure the safety of maintenance personnel, all procedures related to the removal and replacement of these segments, as well as the extraction and handling of their internal components for refurbishment, are conducted with robust protective measures. These measures include the use of specifically designed shielding systems that are tailored to the radiological conditions unique to each operation.

Key maintenance activities—such as disconnecting the vacuum vessels of the segments from adjoining modules, transporting segments from the stellarator hall to the designated hot cell, and relocating internal components to interim storage—are all performed using remotely operated or robotic equipment. This approach minimizes the need for direct human involvement in high-radiation areas, significantly reducing occupational exposure.

When manual intervention is required, such tasks are only permitted if optimized shielding is in place to ensure worker safety. In addition, supplemental shielding is deployed during segment operations to safeguard adjacent plant equipment, thereby maintaining an overall safe environment for both personnel and facility infrastructure.

To effectively mitigate the risks associated with loose contamination - particularly activated dust - within the stellarator, the facility employs a comprehensive removal strategy prior to accessing the unit. Most of the particulate matter is removed through gas flushing operations. This primary method is further supported by through-port vacuum extraction as necessary, ensuring that dust is thoroughly eliminated from the interior surfaces of the vacuum vessel and related components.

During these procedures, any dust that becomes mobilized is promptly captured in disposable filters. The handling and management of these filters are strictly governed by their radioactivity classification. This ensures that all protocols for safe handling and disposal are rigorously followed in accordance with the level of radiological activity present.

Upon opening the stellarator, residual loose contamination is expected to be minimal. These remaining particulates are anticipated to be primarily dense particles originating from tungsten coatings or low-activation steels. Larger particulates will settle onto adjacent surfaces under the influence of gravity. This natural settling process enables targeted retrieval and collection of the dust as it exits the vacuum vessel, supporting the overall safety and radiological protection measures in place during maintenance activities.

The plant's design and maintenance protocols incorporate specific measures to facilitate the safe collection of such contaminants. These provisions extend to both the planned locations where contamination is likely to settle and adjacent areas, ensuring comprehensive management of loose contamination throughout maintenance activities.

Comprehensive radiation and tritium monitoring throughout the facility is fundamental for accurately determining both the location and risk level of radiological sources during maintenance activities. This vital information establishes the basis for effective radiological protection programs.

Once the location and intensity of radiological sources are identified, the core principles of Time, Distance, and Shielding (T.D.S) are implemented to protect personnel against external radiation exposure. The application of these principles includes:

- Limiting the amount of time personnel spend near radiation sources.

- Maximizing the distance between workers and radiation sources wherever possible.

- Utilizing optimized shielding solutions to protect workers from all forms of radiation hazards, ensuring compliance with dose constraints and regulatory limits while following alara (As Low As Reasonably Achievable) guidelines.

These foundational principles guide the development and implementation of risk-appropriate radiation protection strategies and measures. The approach is informed by data collected through advanced radiation safety detection, monitoring, and control systems.

The systematic characterization and assessment of threat levels play a pivotal role in refining operational plans and may involve the following actions:

- Recommending respiratory protection when airborne threats are present.

- Establishing appropriate dwell times, area monitoring protocols (such as rate alarms), and dosimetry requirements.

• Identifying restricted areas or designated waiting zones between activities.

- Specifying additional personal protective equipment (P.P.E) requirements, including those for non-radiological hazards based on the nature of identified risks.

A comprehensive and systematically documented maintenance program discussed early and iteratively with the regulator as design progress is essential. Such a program not only minimizes plant downtime through effective maintenance strategies but also ensures the protection of personnel, equipment, and the public from radiological hazards within any fusion machine.

6. Conclusion

The regulatory landscape is maturing and taking shape at planned pacing in North America, with the United States ^{} , and the United Kingdom, which have made decisions and are regulating fusion technologies with an goal-setting approach, that is, by directing applicants toward the fundamentals and basic principles of radiation safety rather than relying on regulations and radiation safety analysis and assessment rules that are rooted in phenomenological realities and the radiation risks inherent to N.P.P's.

Such a goal-oriented approach directs license applicants toward a radiation safety analysis that reflects the phenomenological realities of different emerging fusion energy technologies, allowing applicants and regulators to focus their efforts on radiation safety issues specific to fusion technologies and avoid potentially excessive debate on exemptions in the context of a nuclear power plant regulatory regime that has evolved considerably and become increasingly specialized and refined through operational experience gained over the past six or seven decades.

The goal-setting approach for an emerging fully integrated fusion energy producing machine enables a regulatory regime evolving in harmony and contributing to the stages of increasing maturity of fusion technologies by focusing on the essentials in terms of safety rules and, like the industry, gaining valuable operational experience to refine the rules gradually. The same applies to the development of fusion-specific codes and standards. It took the nuclear power industry a decade or more to come up with its first comprehensive set of codes and standards in the light of the precious operating experience gained. Some elements of those codes and standards can be adopted after validation through test beds (effects tests) and calibration, but many others won't be suitable given that safety constraints associated with systems included in fusion machines of various technology families greatly vary in purpose, components and phenomena compared to N.P.P's' realities.

With a goal-setting approach anchored in fundamental principles of radiation safety, licensing requirements and codes and standards for fusion technologies can progress through stages of maturity with full and transparent cooperation between industry, regulatory bodies and technical supporting organizations from academia and R&D organizations.

On the other hand, fusion technologies share similarities with other novel nuclear applications (industrial, commercial and medical) when it comes to implementing the fundamental rules and principles of radiation safety analysis and assessment applicable to safety-relevant S.S.C's or addressing issues such as modeling the release limits of radionuclides or any other non-radiological contaminants into the atmosphere or in the form of liquid effluents.

On the former, this paper presented a practical example with the use of the industry mainstream Master Logic Diagram (M.L.D) and Functional Failure Mode Effect Analysis (F.F.M.E.A) approaches to identify radiation safety constraints for safety relevant S.S.C's to inform engineering design integrated requirements from the earliest phases of design for all radiation safety relevant fusion machine S.S.C's.

On the latter, the C.S.A approach for calculating derived tritium release limits provides a systematic and defensible means of verifying compliance with public dose limits, a method that has been used by candu facilities around the world for several decades. It has also been calibrated and validated by multiple efforts to monitor and track tritium levels in communities surrounding the plants. This approach is relevant for calculating the tritium release limits for T.1.E S.F.P.P from the initial design phase to determine the optimal sitting options and inform the performance and reliability requirements for the tritium fuel cycle S.S.C's and all equipment involved in the detection, monitoring, and recovery of parasitic tritium losses in the working environment.

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