Preamble

Molten Salt Tritium Breeding Materials in Fusion Reactors: A Neutronic Comparative Analysis for ITER
Alper Karakoç

Abstract

This study investigates the effects of varying tritium breeding materials and their lithium enrichment rates on the Tritium Breeding Ratio (TBR) and Energy Multiplication Factor (M) within the tritium breeding zone of a fusion reactor. The magnetic fusion reactor model was developed based on the geometric and plasma parameters of the International Thermonuclear Experimental Reactor (ITER), ensuring a realistic representation of current fusion reactor designs. ITER-grade stainless steel (SS 316 LN-IG) was selected as the first wall material due to its excellent mechanical properties, high resistance to radiation damage, and compatibility with high-temperature environments. The coolant and tritium breeding materials considered in the blanket included natural lithium, lithium fluoride (LiF), lithium nitride (Li₃N), FLiBe (LiF-BeF₂), and FLiNaBe (LiF-NaF-BeF₂). These materials were chosen for their ability to facilitate tritium breeding while maintaining thermal and neutronic efficiency. Neutron transport calculations and geometric modeling were performed using the widely recognized 3D simulation tools MCNP5 and TopMC, which employ the continuous-energy Monte Carlo method. The simulations utilized built-in continuous-energy nuclear and atomic data libraries, along with the Evaluated Nuclear Data File (ENDF) system (ENDF/B-V and ENDF/B-VI), ensuring reliable and validated results.

The results highlight the importance of material selection and enrichment optimization in achieving efficient tritium breeding and energy production. FLiBe, in particular, shows promise for future fusion reactor designs due to its superior performance in terms of TBR and M. These findings provide valuable insights for the development of sustainable and high-performance fusion reactors, contributing to the global pursuit of clean and virtually limitless energy.

Key Words: ITER, Magnetic Fusion Reactor, Tritium Breeding Ratio, Energy Multiplication Factor, Neutronic Analysis

1. Introduction

In the contemporary era, sustainable technological advancements and meeting humanity's ever-increasing demands are intrinsically linked to energy production and utilization. The exponential growth of human populations globally, together with processes of industrialization and urbanization, has resulted in an unprecedented rise in energy demands, making energy a pillar of economic and social development. Researchers and scientists are intensively exploring both conventional and alternative energy supplies to cope with this challenge. Among these options, nuclear power represents one of the most important solutions, offering a sustainable, low-carbon, and highly efficient method of energy production. The last century has witnessed significant scientific efforts in developing nuclear energy technologies, which play a crucial role in meeting global energy requirements while minimizing environmental impacts, thereby setting the stage for advancements in fusion energy.

Compared to renewable energy production methods, nuclear energy provides a consistent and substantial amount of energy throughout the day, ensuring a stable supply for consumers. Relative to fossil fuel-powered plants, nuclear energy uses significantly less fuel, consumes fewer materials, and produces zero carbon emissions, making it a sustainable energy source. Today's nuclear power plants utilize fission reactions, in which energy is generated by splitting atomic nuclei. However, a significant factor limiting the sustainability of fission reactors is the limited availability of fuel reserves, such as uranium and plutonium. In addition, both short-lived and long-lived radioactive waste are generated during the operation of fission reactors; if these wastes are not managed and stored according to established protocols, they can lead to serious environmental consequences. Beyond the necessity of storing these wastes and the associated environmental risks, high fuel supply and procurement costs significantly reduce the advantages of fission reactors. Unlike fission reactors, fusion reactors have significant potential to meet the need for sustainable nuclear energy because the fuel used in these reactors is almost unlimited. The abundance and easy accessibility of fusion fuel, as well as the absence of risks such as reactor meltdown or loss of control, have encouraged scientists and engineers to conduct extensive research on the feasibility of fusion technology. Studies conducted in the 1920s investigated the fundamental principles of fusion reactions and demonstrated that the energy generated in stars originates from these reactions [1]. This theory was later strengthened by mathematical models developed by Robert D'Escourt Atkinson and Fritz Houtermans, which revealed that the energy source in stars is nuclear fusion [2]. In addition, these studies demonstrated that fusion reactions are not only dependent on extremely high temperatures in stars but can also be carried out on Earth. These early studies provided significant momentum for the feasibility and development of fusion reactors. In 1950, Andrei Sakharov and Igor Yevgenyevich Tamm proposed the concept of TOKAMAK (Toroidal Camera and Magnetic Coil), a type of magnetic confinement fusion device [3]. Shortly afterwards, in 1951, Lyman Spitzer developed the Stellarator device, while Richard F. Post and Gersh Budker independently presented the concept of the magnetic mirror [4]. The TOKAMAK concept emerged as a significant breakthrough in magnetic confinement systems in the late 1960s thanks to experimental studies conducted by Lev Artsimovich. Following this development, TOKAMAK gained priority in research and development as the most suitable approach for achieving controlled nuclear fusion. Today, magnetic confinement fusion continues to be the dominant methodology in fusion reactor designs, with intensive research being conducted on three basic configurations: TOKAMAK, Stellarator, and Magnetic Mirror systems. In the early 1980s, comparative experiments were conducted between TOKAMAK and Magnetic Mirror configurations at the Princeton Plasma Physics Laboratory, and the technological feasibility of these systems was evaluated. As a result of these studies, it was understood that the TOKAMAK configuration was much more compatible with the operational requirements of fusion reactor technologies, and this design emerged as the preferred system [5].

The International Thermonuclear Experimental Reactor (ITER), which builds on decades of advances in magnetic confinement physics and plasma stability, represents the most advanced implementation of the tokamak concept [6]. ITER, the world's largest fusion experiment, aims to demonstrate the feasibility of deuterium-tritium (D-T) fusion as a sustainable energy source, targeting a tenfold energy gain (Q ≥ 10) through improved plasma confinement [7]. The design of ITER includes advances in materials engineering and superconducting magnet technology, with a primary focus on achieving sustained fusion combustion, a vital step toward realizing practical fusion energy [8]. The ITER Project has continued to progress despite geometrical incompatibilities and corrosion issues in some components. A new project plan has been determined, system installations have been completed while repairs are ongoing, and most major components have been delivered. Initial experimental work is planned to begin in 2034, with the deuterium-tritium phase postponed to 2039 [9]. Developed through international collaboration, ITER builds on the knowledge gained from previous tokamak experiments such as the Joint European Torus (JET), which broke records in plasma performance [10], and the Experimental Advanced Superconducting Tokamak (EAST), which has contributed to steady-state high-confinement plasma research [11].

One of the main challenges for deuterium-tritium (D-T) fueled reactors is the ability to produce tritium in a self-sufficient manner. Since tritium is both a rare and radioactive isotope, it must be produced continuously in the breeding blanket of the reactor to sustain the fusion process. This production relies on neutron-lithium interactions within the blanket and requires carefully designed systems that optimize neutron utilization, thermal management, and material performance [12]. The success of this production process has a direct impact on the reactor's long-term operating capability and its potential as a commercially sustainable energy source. To overcome the challenges of tritium production capability, extensive research is being conducted to optimize tritium breeding materials (TBM) and maximize the tritium breeding ratio (TBR) in fusion reactor blankets. Liquid tritium breeders such as lithium-lead (LiPb) have traditionally been preferred due to their high tritium production potential. However, concerns about lead-producing polonium due to neutron activation have increased interest in lead-free alternatives such as molten salts and solid breeders [13-14-15].

Among these options, lithium fluoride-beryllium fluoride (FLiBe) and lithium fluoride (LiF) stand out due to their compatibility with high neutron flux environments, low long-term activation levels, and lack of polonium-related safety risks [16-17]. Natural lithium (Li) and lithium nitride (Li₃N) have also been investigated due to their high lithium content and relatively simple tritium extraction methods. However, challenges such as neutron moderation efficiency, thermal-hydraulic properties, and chemical stability under irradiation remain the focus of research [18-19-20]. Recent neutronic simulations have shown that FLiBe-based blankets can achieve tritium production rates (TBR) above 1.1 in single-fluid configurations, especially when combined with beryllium neutron multipliers. This rate meets the minimum required for fuel self-sufficiency [21]. Studies conducted to ensure tritium self-sufficiency in fusion reactors have shown that the TBR value must exceed at least 1.05 [22]. To ensure sufficient tritium production, the effect of the thickness and density of the first wall material must be evaluated in reactor designs. Meeting this requirement necessitates optimizing lithium enrichment and improving the blanket design to increase neutron capture and tritium production efficiency. Another effective way to increase tritium production is lithium enrichment. This method increases the density of the ⁶Li isotope in the breeding materials, thus increasing both the TBR and providing a more efficient and sustainable fuel cycle for fusion reactors. Tritium production studies conducted for the DEMO fusion reactor determined that increasing the ⁶Li isotope density in the tritium production zone is a key strategy [24].

After ensuring sustainable tritium production in fusion reactors, design optimizations should be made to ensure that the energy produced in the reactor exceeds the energy consumed. In this context, the reactor's blanket structure should have a high ability to capture neutron energy and convert this energy into a useful form. This conversion ability is expressed by the energy multiplication factor (M), and studies in the literature have examined the variation of the M value across different reactor concepts. In a study conducted using the Helium-Cooled Lithium Lead Blanket concept in the DEMO fusion reactor, this value was calculated to be 1.17 [25]. In another study using the Helium-Cooled Pebble Bed Blanket concept, the energy multiplication factor was calculated to be 1.22 [26].

In addition to ensuring tritium self-sufficiency and maximizing the energy multiplication factor, the selection of durable materials for the first wall structure is also critical for the success of fusion reactors. Among the available options, austenitic stainless steel SS 316 LN-IG (ITER grade) stands out due to its superior mechanical strength, radiation resistance, and ability to withstand the extreme conditions of fusion environments. This material, which is also widely used in ITER, is preferred for the first wall structure due to its stability in high-temperature conditions and its reduction of adverse effects such as swelling and embrittlement that may occur under radiation. The first wall, which is directly exposed to the plasma, must withstand harsh conditions such as extreme thermal loads, neutron radiation, and particle bombardment; therefore, material selection is crucial for the durability and safety of the reactor. SS 316 LN-IG is a low-carbon and nitrogen-strengthened 316 stainless steel type designed to withstand these harsh conditions. Thanks to its reduction of problems such as swelling and embrittlement due to radiation, its excellent weldability, and high fracture toughness, it is highly suitable for large-scale fusion applications such as ITER's blanket modules and divertor components [27].

2.1. Description of Magnetic Fusion Reactor Geometry

Figure 1 [FIGURE:1] shows the three-dimensional cross-sectional geometry and main components of the magnetic fusion reactor. The geometry was created using the TopMC program, taking reference from the ITER design parameters. The model includes essential reactor parts such as the vacuum vessel, blanket, divertor, toroidal and poloidal field coils, central solenoid, thermal shield, and tritium breeding modules.

Figure 1. 3D Cross-sectional view of magnetic fusion reactor

The materials and functional roles of these main reactor components are summarized in Table 1 [TABLE:1]. This table provides an overview of the selected materials and their corresponding application areas in the modeled fusion reactor based on ITER design principles. Figures 2 [FIGURE:2] and 3 [FIGURE:3] illustrate poloidal and toroidal cross-sectional views of the magnetic fusion reactor modeled in MCNP5.

Table 1. Materials and Functions of Modeled Fusion Reactor Components Based on ITER Design

Component Material Application Area / Function Plasma • Deuterium (50%)
• Tritium (50%) D-T plasma is the main fusion fuel, enabling energy-releasing reactions at high temperatures and densities Shielding & First Wall • SS 316L(N)-IG
• Tungsten • Plasma facing and shielding component
• Plasma-facing surface (strike points) Divertor • CuCrZr
• SS 316L(N)-IG • Heat sink
• Structural support and manifold structures Tritium Breeding Material & Coolant • Li (natural)
• LiF
• Li₃N
• FLiBe
• FLiNaBe Serve as tritium breeders and coolants, enhancing tritium production and thermal management in reactor blanket Neutron Multiplier Be₁₂Ti Primary neutron multiplier in blanket design Reflector Graphite Enhances tritium production by reflecting and slowing down escaping neutrons back into the blanket and fuel region Vacuum Vessel • SS 316L(N)-IG
• B₄C
• Inconel 625 / 718 • Main structural material (vessel body)
• Neutron shielding (in localized regions)
• Support elements Thermal Shield SS 316L(N)-IG Reduces thermal load on cryogenic magnets by absorbing and reradiating plasma heat Central Solenoid Nb₃Sn Generates central loop voltage to initiate and sustain plasma current Toroidal Field Coil Nb₃Sn Main conductor for toroidal field coils Poloidal Field Coil NbTi Main conductor for poloidal field coils

Figure 2. Poloidal cross-sectional view of magnetic fusion reactor

Figure 3. Toroidal cross-sectional view of magnetic fusion reactor

In these models, the plasma region is defined using ITER's deuterium-tritium (D-T) plasma parameters, featuring a low aspect ratio with a major radius of 6.2 m and a minor radius of 2 m. The plasma chamber is enclosed by a layered structure comprising, from innermost to outermost: the first wall, divertor, tritium breeding zone, neutron multiplier zone, reflector zone, thermal shield, vacuum vessel, and magnetic coils. Figure 4 [FIGURE:4] shows the thickness of the blanket structure layers of the modeled reactor.

Figure 4. Reactor structure with radial distance

2.1.1. Plasma Chamber

The plasma chamber serves as the confinement region for the plasma generated by deuterium-tritium (D-T) fusion reactions. High-energy neutrons produced in the plasma region during these reactions interact with the first wall. In the model, the neutron source is represented based on ITER's plasma parameters.

2.1.2. First Wall

The first wall is the interior surface of the tokamak. It lies closest to the plasma, acting as the initial barrier that faces intense heat and particle flux generated during fusion reactions. Therefore, selecting appropriate materials for the first wall is critical to ensure structural integrity and performance under extreme conditions. The first wall material must resist high temperatures and intense neutron flux. Additionally, it should possess excellent hardness, corrosion resistance, and a low neutron absorption cross-section to minimize adverse effects on the reactor's neutron economy. This study adopted a solid first wall design, utilizing SS 316 LN-IG as the structural material due to its superior mechanical and neutronic properties.

2.1.3. Divertor

The divertor is a vital component in fusion reactors, particularly in TOKAMAK-type systems. Its primary functions include maintaining plasma purity by preventing contact with the blanket structure and evacuating fuel particles from the plasma. These roles are essential for ensuring the efficient and safe operation of fusion reactors.

2.1.4. Tritium Breeding Zone

Fusion reactors employ a dual-coolant system designed to facilitate both efficient heat transfer and tritium production. The primary coolant absorbs heat generated within the reactor and transfers it to a secondary coolant loop for energy conversion. At the same time, it serves as the medium for tritium breeding, where lithium isotopes (⁶Li and ⁷Li) undergo neutron-induced reactions (n, α) and (n, nα) to generate tritium (³H). The selection of the primary coolant is a key design consideration and must meet several essential criteria:

• High lithium density to maximize tritium production by ensuring a sufficient supply of lithium for neutron interactions
• High specific heat and thermal conductivity to enable effective heat removal and maintain uniform temperature distribution
• Low density and viscosity to improve coolant flow dynamics and minimize the energy required for circulation
• Minimal corrosion and erosion effects to enhance the durability of structural components exposed to intense neutron flux
• A high boiling point and low melting point to ensure stable operation under extreme thermal conditions
• A low neutron absorption cross-section to optimize neutron economy and enhance tritium breeding efficiency
• Cost-effectiveness to support the feasibility of large-scale reactor deployment

In this study, molten salts—natural lithium (Li), lithium fluoride (LiF), lithium nitride (Li₃N), FLiBe (LiF-BeF₂), and FLiNaBe (LiF-NaF-BeF₂)—were evaluated as primary coolant candidates due to their advantageous thermal and neutronic characteristics [28]. Their high lithium content, excellent thermal stability, and low neutron absorption make them well-suited for tritium breeding. Furthermore, their compatibility with structural materials and ability to sustain high operating temperatures strengthen their potential for use in magnetic fusion reactor systems.

2.1.5. Neutron Multiplier Zone

The neutron multiplier zone is a critical component for enhancing tritium production, improving neutron economy, and ensuring a sustainable fuel cycle in fusion reactors. This zone amplifies the number of neutrons generated during D-T fusion reactions, thereby optimizing tritium breeding. In this study, Be₁₂Ti was selected as the neutron multiplier material due to its favorable neutronic properties [29].

2.1.6. Reflector

The reflector layer redirects neutrons escaping from the coolant zone back into the reactor, increasing the likelihood of neutron interactions within the molten salt-fuel mixture. This layer minimizes neutron leakage and supports tritium breeding. In this study, graphite, which exhibits minimal neutron absorption, was used as the reflector material.

2.1.7. Thermal Shield

The thermal shield is designed to protect the ultra-cold superconducting magnets from excessive nuclear heating, radiation damage, and neutron flux. It also ensures that radiation exposure to diagnostic tools and maintenance equipment remains within acceptable limits. The shield mitigates temperature variations, as superconducting magnets in TOKAMAK reactors operate at cryogenic temperatures (4 K). Each watt of thermal energy deposited in the magnets by neutrons and gamma rays requires approximately 500 watts of cooling energy to maintain operational stability [30].

2.1.8. Vacuum Vessel

The vacuum vessel is a confinement barrier in magnetic fusion reactor designs, ensuring minimal heat accumulation in specific regions, especially magnetic field coils. By eliminating isotopes in these areas, heat transfer via convection and conduction is prevented. In the modeled reactor, a vacuum layer was incorporated between the insulation layer and the magnets to protect the magnets from heat-induced degradation.

2.1.9. Magnetic Coils

Superconducting magnets are employed in magnetic fusion reactors to confine plasma within the reactor. Positioned in the outermost layer, these magnets generate magnetic fields that prevent plasma dispersion. Each type of magnet in the reactor serves a distinct purpose within the overall system: a central solenoid functioning as a large transformer to induce and sustain strong plasma currents during extended pulses; a set of six horizontal poloidal field coils positioned outside the toroidal magnet structure to control plasma shape and stability; and eighteen D-shaped vertical toroidal field coils surrounding the vacuum vessel, creating a magnetic bottle for plasma confinement. In this study, Nb₃Sn and NbTi were selected as magnet materials based on the specific magnetic field requirements of each coil system.

3. Simulation Tools

In this study, the simulation tools MCNP and TopMC were used sequentially. The geometry designed in MCNP was tested in the TopMC program, and after making the necessary adjustments, it was recreated as an MCNP input file. This generated input file was used as a common input file for both MCNP and TopMC, and the simulation results were verified by ensuring the consistency of the output files produced by both programs.

3.1. MCNP

MCNP (Monte Carlo N-Particle) is a program used to model the transport of neutrons, photons, electrons, and other particles, performing simulations with the Monte Carlo technique. It is widely applied in nuclear engineering, radiation protection, and various fields requiring particle transport analysis. The code is developed and maintained by Los Alamos National Laboratory (LANL) in the United States. The fifth version, MCNP5, leverages the Monte Carlo method, a statistical technique that simulates the behavior of particles by following their individual trajectories through a defined geometry [31]. This method is particularly beneficial for simulating complex systems and is advantageous in neutron transport problems, where deterministic approaches may struggle with intricate geometries and material interactions.

MCNP5 employs a continuous-energy Monte Carlo method, which means it doesn't rely on predefined energy groups. Instead, it tracks particles across a seamless range of energies. This is especially important for accurately modeling neutron behavior, as neutrons often span a wide energy spectrum in many nuclear and radiation-related applications. Additionally, MCNP5's three-dimensional modeling capabilities enable the simulation of complex geometries with great detail, providing users with the flexibility to accurately represent intricate systems. This makes MCNP5 a valuable resource for researchers and engineers who apply it across a wide spectrum of fields, including reactor physics, radiation shielding, medical physics, and environmental radiation transport studies. Overall, MCNP5 serves as a robust tool for predicting and analyzing particle interactions in various environments.

3.2. TopMC

TopMC (Multifunctional Program for Neutronic Computing, Nuclear Design and Safety Assessment) is an advanced nuclear design software developed by the FDS (Fusion Design and Simulation) Consortium for over thirty years. TopMC, an improved and extended version of SuperMC, serves as a large-scale tool for performing comprehensive neutronic calculations. Its main function is radiation transport calculations covering all neutronic processes such as depletion, radiation source term analysis, dose assessment, biohazard analysis, material activation and transmutation. TopMC stands out with its high efficiency, precision and multi-physics capabilities; it offers accurate analytical modeling and visualization tools [32]. It also provides virtual simulation features, intelligent nuclear design and powerful safety assessment functions. Although this software is mainly used in the design and safety analysis of nuclear energy systems, it also finds application in various fields of nuclear technology such as radiation medicine and nuclear detection.

4.1. Tritium Breeding Ratio

The tritium breeding ratio is defined as the ratio of tritium produced in the blanket structure of a fusion reactor to the tritium consumed in the plasma. Tritium, a key isotope for deuterium-tritium (D-T) fusion reactions, is not naturally abundant and must be continuously bred within the reactor's blanket to sustain the fusion process. Due to its radioactive instability and short half-life (approximately 12.3 years), tritium must be utilized shortly after production. If stored as a reserve, the required TBR for reactor operation increases over time to account for decay and other losses. To ensure sufficient tritium production and sustainable fusion reactions, studies indicate that the TBR must be at least 1.05. Tritium production in magnetic fusion nuclear reactors relies on the complex interactions between thermal and fast neutrons and lithium isotopes. Lithium exists in two primary isotopes: ⁶Li (7.6% natural abundance) and ⁷Li (92.4% natural abundance). Tritium generation is governed by two key nuclear reactions: the exothermic ⁶Li(n, t) reaction, where thermal neutrons interact with ⁶Li to produce tritium, and the endothermic ⁷Li(n, t) reaction, where fast neutrons interact with ⁷Li to yield tritium. These reactions are fundamental to maintaining a sustainable tritium supply, which is essential for fueling fusion reactions in magnetic fusion reactors. The ⁶Li isotope exhibits a significantly higher propensity for (n, t) reactions with thermal neutrons compared to ⁷Li. Consequently, ⁶Li plays a pivotal role in determining the TBR. The enrichment ratio of ⁶Li and the thickness of the coolant layer are critical factors influencing the TBR, as they directly affect neutron capture efficiency and tritium production rate. The TBR results obtained as a function of ⁶Li enrichment for various materials are shown in Figure 5 [FIGURE:5].

Figure 5. TBR value with various tritium breeding material and lithium enrichment

4.2. Energy Multiplication Factor

The energy multiplication factor (M) is a key parameter used to evaluate the energy performance of the fusion reactor blanket system. It is defined as the ratio of the total energy stored or recovered in the system to the kinetic energy of the incident fusion neutrons. During fusion reactions, not all the released energy is retained in the system; energy is lost through escaping neutrons, alpha particles, and gamma radiation. Since neutrons are uncharged, they travel beyond the plasma and deposit their energy in surrounding materials. To achieve net thermal power generation, the energy recovered from these reactions must exceed the energy initially carried by the plasma, making a sufficiently high energy multiplication factor essential. For fusion systems based on D-T fuel, a minimum M value of 1.2 is generally considered necessary to offset system losses and contribute meaningfully to thermal energy output [33].

A well-designed blanket system must optimize these reactions to balance both tritium production and thermal energy recovery. Additionally, materials like beryllium, used as a neutron multiplier, can improve neutron economy and enhance both the energy multiplication factor and the tritium breeding ratio. Geometry, material selection, and neutron transport characteristics all influence how effectively the reactor can utilize fusion neutron energy. In calculating the energy multiplication factor, the two most significant factors in terms of energy production and consumption are the exothermic ⁶Li(n, α)T and endothermic ⁷Li(n, αn)T reaction energies. Therefore, the energy multiplication factor is calculated as follows:

$$M = 1 + 4.784 \times T_6 - 2.467 \times T_7$$

In this equation, the values T₆ and T₇ represent the tritium production rates of the ⁶Li and ⁷Li isotopes, respectively. The energy multiplication factor results from this study are shown in Figure 6 [FIGURE:6].

Figure 6. M value with various tritium breeding material and lithium enrichment

5. Conclusion

In this study, detailed neutronic simulations were performed to assess the influence of different tritium breeding materials as well as varying levels of lithium enrichment on the overall performance of the fusion reactor blanket system. Considering the geometrical and operational constraints, the calculations were carried out with a three-dimensional code for a system modeled on the ITER reference design. The input file created in the TopMC program was executed in both MCNP5 and TopMC, and the consistency of the results confirmed the reliability of the neutronic calculations in this study. The blanket layout was studied under steady-state plasma conditions, with special attention to important performance parameters such as the TBR and M.

The simulation results show that tritium production increases when lithium enrichment is raised to an optimum level, followed by a decrease. This non-linear behavior can be attributed to the isotopic structure of lithium. While increasing enrichment raises the quantity of ⁶Li, which can generate tritium directly through the exothermic ⁶Li(n,α)T reaction, it also reduces the loading of ⁷Li, which serves as a secondary neutron moderator and contributes to neutron economy by producing tritium via the ⁷Li(n, n′α)T reaction. The depletion of ⁷Li at low levels causes a decrease in neutron moderation, affecting the thermal neutrons needed to sustain tritium breeding reactions. In contrast, the energy multiplication factor shows a consistent upward trend with increasing ⁶Li content, as the reaction cross-section of ⁶Li(n, α)T is significantly higher at lower neutron energies.

Upon analysis of the results, FLiNaBe and LiF were found to exhibit sub-threshold neutronic performance with respect to the required TBR limit. Although FLiBe succeeded in exceeding the tritium breeding ratio value within a moderate enrichment range (20–60%), it fell short of the requirement at lower and higher enrichment levels, indicating sensitivity to lithium enrichment. Natural lithium and lithium nitride (Li₃N) also did not meet the required threshold, with calculated tritium breeding ratio values remaining below target even at 90% and 50% enrichment, respectively. These results are largely affected by their unfavorable moderation characteristics and lower neutron economy compared to enriched FLiBe mixtures. In terms of energy multiplication, FLiBe consistently delivered the highest energy multiplication factor values across the enrichment spectrum, especially in configurations where the tritium breeding ratio remained above the minimum threshold. This demonstrates FLiBe's dual effectiveness in enhancing both tritium production and thermal energy generation.

As a result, according to the simulation results based on ITER geometry and plasma input parameters, FLiBe appears to be the most suitable tritium breeding material for a future magnetic fusion reactor concept. Its ability to support self-sustaining fuel cycles while delivering higher thermal output makes it a strong candidate for integration into environmentally sustainable and economically viable magnetic fusion systems. The findings from this analysis, particularly regarding the role of FLiBe, are expected to inform material selection strategies and blanket design optimizations in upcoming fusion energy projects.

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