Breeding Ratio Research Articles

Impact of trapping on tritium self-sufficiency and tritium inventories in fusion power plant fuel cycles

Abstract The dynamic analysis of fusion power plant (FPP) fuel cycles highlights the challenge of achieving tritium self-sufficiency in future FPPs. While state-of-the-art fuel cycle models offer valuable insights into the necessary design parameters for attaining tritium self-sufficiency, none of these models currently consider the impact of tritium trapping within fuel cycle components. However, detailed analysis of individual components reveals that substantial amounts of tritium can be trapped within the first wall, divertors, and breeding blanket systems, suggesting that tritium trapping may significantly influence the FPP ability to achieve self-sufficiency. The compounded effects of additional tritium traps generated by irradiation effects and component replacements further exacerbate this challenge. The novelty of this work is the integration of an explicit, physics-based model for tritium trapping, evolution of damage-induced traps, and component replacements into a dynamic, system-level model of a fuel cycle. The results show an increase of a factor 103 - 104 of tritium inventory in the first wall and vacuum vessel of an ARC-class FPP when accounting for the aforementioned phenomena. This, coupled with the replacement of components subject to significant tritium trapping, slows down fuel cycle dynamics, resulting in an extended tritium doubling time (50% increase), higher start-up inventory (30% increase), and higher required tritium breeding ratio (2-5%) compared to a scenario without tritium trapping.

Development of fuel depletion code for molten salt reactor with very deep burnup

A liquid-fueled molten salt reactor (MSR) can reach a deep burnup based on online reprocessing and continuously refueling, which requires significantly different burnup calculation methods for MSRs compared with those for the traditional reactors. To address the unique burnup features and consider the fidelity of isotopic evolution in an MSR, a fuel depletion code ThorMCB is developed based on the OpenMC coupled with a specific depletion code, MODEC. Furthermore, to lower the computational cost of acquiring the equilibrium state through the time evolution step by step for an MSR, an equilibrium burnup calculation code ThorMCB-eq based on the OpenMC and MODEC is developed, which can obtain the equilibrium burnup efficiently. A single fuel lattice of MSR and an a Molten Salt Fast Reactor (MSFR) benchmark are applied for verifying the correctness of the ThorMCB and ThorMCB-eq codes. Compared with a neutron transport calculation code KENO-VI coupled with MODEC, the maximum deviation of the dominant heavy nuclides (HNs) at equilibrium state by ThorMCB is less than 10%, and that of the total mass of fission products (FPs) is less than 3%. For the MSFR benchmark, the neutronic parameters including temperature reactivity coefficient, the mass evolution of main HNs and FPs and breeding ratio (BR) from ThorMCB agree with the references. The equilibrium behavior can be quickly obtained with ThorMCB-eq, and the relative mass deviations of most nuclides keep around 2% in comparison with the results of step-by-step burnup evolution with ThorMCB. Furthermore, the same fuel contents and micro one-group cross sections at equilibrium are obtained with two different types of start-up fuels and a constant power density and fuel reprocessing scheme. In conclusion, the verified results indicate that ThorMCB and ThorMCB-eq can both provide reliable simulation for depletion evolution and equilibrium burnup for MSR fuel cycle.

Pareto front of sodium void worth and breeding ratio in metal-fueled sodium fast reactor with axially heterogeneous core design by coupling neutronics and genetic algorithm

For sodium-cooled fast reactors (SFRs), achieving low sodium void worth (SVW) and flexible breeding ratios (BR) is crucial. This study establishes a method coupling neutronics simulations with genetic algorithms (GA) and applies it to a 3000 MWth metal-fueled SFR to optimize the SVW, BR, and the power distribution of axially heterogeneous cores. The Pareto front, i.e. optimal front, between SVW and BR is determined, quantitatively clarifying the trade-off relationship between these two parameters. Axially heterogeneous cores with equal inner and outer fissile heights can achieve SVW adjustments from −71 to 3148 pcm and BR adjustments from 1.26 to 1.62. Designs with unequal fissile heights slightly expand the feasible region, but the impact on the optimal front for equal-height designs is limited. Through the analysis of the spatial distribution of the sodium void effect, it is shown that the bond sodium model significantly influences the SVW in low sodium void cores, while its impact on high breeding cores is minimal. The results indicate that the power distribution can be optimized by adjusting the enrichment of subregions within the constraints of SVW and BR. For low sodium void cores, retaining a certain amount of lower fertile material is necessary to flatten the power distribution. While the criticality search during depletion calculations affects the parameters, its impact on identifying the optimal front is limited.

Optimization of Tokamak Blanket Design to Minimize Neutron Flux on Magnets

Fusion energy, a promising solution to global energy challenges, replicates the same processes that give our sun its energy, notably through deuterium-tritium (D-T) fusion reactions that produce significant energy. This study explores the mechanics and challenges of various fusion reactions, emphasizing the pivotal role of lithium breeder blankets in producing tritium, essential for sustaining D-T fusion in tokamak reactors. The helium-cooled lithium lead (LiPb) blanket design was simulated to optimize tritium breeding and neutron flux management. Results indicated a tritium breeding ratio (TBR) of 1.15, surpassing the self-sufficiency target of 1.1, with further improvements through increased lithium content and blanket thickness. Effective neutron shielding ensured safe operational limits for reactor components. These findings demonstrate the feasibility of achieving self-sustaining fusion reactions, essential for the viability of fusion power as a sustainable energy source. Future research will focus on advanced materials, refined simulations, and enhanced cooling technologies to further optimize fusion reactor designs.

Fusing together an outline design for sustained fuelling and tritium self-sufficiency.

Ensuring tritium fuel self-sufficiency while maintaining continuous and high-specification fuel flow to the tokamak via a low tritium inventory and controllable fuel cycle is a significant challenge to the STEP plant design. Effective and high-quality fuelling and exhaust design is required to sustain and control a stable plasma, whereas fuel sufficiency is required to prevent depletion of available tritium supply. Concerns regarding the lack of tritium availability preventing continuous tritium import are countered by breeding, where highly energetic neutrons from the core fusion reactions interact with lithium atoms suspended in the surrounding breeder blanket to produce tritium. The compact nature of STEP prohibits the integration of inboard breeder blankets posing a significant challenge for the design team looking to ensure more tritium is bred and made available than consumed within the core plasma. This paper outlines how purposeful technology selection and system architecting has converged on the outline of a conceivable and tritium-capable fuel cycle and breeder blanket design. Before introducing the STEP fuel cycle design outline and summarizing the approach undertaken to address the challenges facing plasma fuelling, key aspects of fuel self-sufficiency are discussed. This includes discussing a proposed helium-cooled liquid lithium breeder blanket and possible technology options for tritium extraction from lithium. Lastly, there is a brief process modelling overview, which emphasizes the central contribution of various employed modelling methods. Reflections on the presented fuel cycle design outline conclude that substantial development work is still required to realize a continuous tritium fuel cycle design and overcome the major challenges posed by tritium and lithium handling. Reflections on the presented breeder blanket design proposal conclude that while many substantial risks and blockers remain to achieve fuel self-sufficiency, high breeding ratios are expected to be achievable with a compact spherical tokamak configuration. Nonetheless, it is recognized that further consideration is required to ensure that the selection of liquid lithium as a breeder medium provides the overall simplest route to a self-sufficient and realizable design.This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

MANTA: a negative-triangularity NASEM-compliant fusion pilot plant

Abstract The MANTA (Modular Adjustable Negative Triangularity ARC-class) design study investigated how negative-triangularity (NT) may be leveraged in a compact, fusion pilot plant (FPP) to take a ‘power-handling first’ approach. The result is a pulsed, radiative, ELM-free tokamak that satisfies and exceeds the FPP requirements described in the 2021 National Academies of Sciences, Engineering, and Medicine (NASEM) report ‘Bringing Fusion to the U.S. Grid’ (2021 Bringing Fusion to the U.S. Grid). A self-consistent integrated modeling workflow predicts a fusion power of 450 MW and a plasma gain of 11.5 with only 23.5 MW of power to the scrape-off layer (SOL). This low P SOL together with impurity seeding and high density at the separatrix results in a peak heat flux of just 2.8 MW m−2. MANTA’s high aspect ratio provides space for a large central solenoid (CS), resulting in ∼15 minute inductive pulses. In spite of the high B fields on the CS and the other REBCO-based magnets, the electromagnetic stresses remain below structural and critical current density limits. Iterative optimization of neutron shielding and tritium breeding blanket yield tritium self-sufficiency with a breeding ratio of 1.15, a blanket power multiplication factor of 1.11, toroidal field coil lifetimes of 3100 ± 400 MW · yr, and poloidal field coil lifetimes of at least 890 ± 40 MW · yr. Following balance of plant modeling, MANTA is projected to generate 90 MW of net electricity at an electricity gain factor of ∼ 2.4 . Systems-level economic analysis estimates an overnight cost of US$3.4 billion, meeting the NASEM FPP requirement that this first-of-a-kind be less than US$5 billion. The toroidal field coil cost and replacement time are the most critical upfront and lifetime cost drivers, respectively.

Pre-conceptual design and proof of principle assessments of self-cooled Toroidally symmetric lead-lithium (TSLL) blanket

A new self-cooled liquid metal blanket concept called TSLL (Toroidally Symmetric Lead-Lithium) blanket is proposed and assessed, including analysis for magnetohydrodynamic (MHD) flows, structural analysis, and heat transfer and neutronics assessments using the ARC reactor with demountable magnets designed by the Commonwealth Fusion Systems (CFS) as a testbed. The proposed blanket utilizes lead-lithium (PbLi) alloy as breeder/coolant and reduced activation ferritic/martensitic (RAFM) steel as structural material. A special feature of the new concept is the toroidally symmetric flow in the blanket integrated first wall and the breeding zone to reduce the MHD pressure drop, while using anchor links to strengthen the first wall construction. Provided analysis suggests acceptable MHD pressure drop, required mechanical integrity and high tritium breeding ratio of ∼ 1.64. As a result of these assessments, the new blanket concept can be recommended for more detailed studies as a promising blanket candidate for implementation in future fusion devices.

Comparative Study on the Reduction of Tritium Breeding Ratio Caused by Inventory Changes of a Solid-State Tritium Breeding Blanket in a Fusion Demonstration Reactor Using MCNP and FISPACT-II

The self-sufficient supply of tritium in the deuterium-tritium nuclear fusion demonstration reactor (DEMO) is a fundamental design requirement. But, it is hindered by depletion of tritium breeding materials resulting in reduction of tritium breeding ratio (TBR) less than the initial design value especially in the solid-state tritium breeding blanket (TBB) of the DEMO. Unlike the liquid tritium breeding blanket of DEMO, compensation measures of the depleted breeding material in the solid-state TBB will be its substitution depending on the reduction rate of TBR. To estimate the replacement period of the solid-state TBB, it is required to estimate the reduction rate of TBR according to the operation conditions of the DEMO and the physical configuration of a solid-state TBB. In this study, the representative simulation codes, MCNP and FISPACT-II, are used for assessment of the reduction rate of TBR with the benchmarking model which is modified from the one poloidal segment of the TBB in the Korean-DEMO. After 3 full power-year operations with the neutron irradiation with the benchmarking model, the TBR simulated by MCNP is reduced to 96.84% of the initially calculated TBR, but the TBR calculated by FISPACT-II is reduced to 90.57% of the initially calculated TBR.

Assessment of hydrogen production potential of APEX fusion blanket via cobalt-chlorine and copper-chlorine cycles

In this paper, the neutronic and hydrogen production performances of the advanced power extraction fusion reactor (APEX) have been investigated by using TRISO thorium fuel. With this study, TRISO-coated nuclear fuel was first used in the APEX fusion reactor. In this calculation, firstly, the neutronic performances for 10 % ThC + 90 % FLiBe and various 6Li enrichment ratios have been computed with MCNP. Secondly, the potential of hydrogen production of the hydrogen unit, which contains cobalt-chlorine cycle and copper-chlorine cycle, integrated with the advanced power extraction fusion reactor has been performed. The tritium breeding ratio has changed between 1.09 and 1.15 among 7.5 % and 90 % 6Li enrichment ratios for 10 % ThC + 90 % FliBe. The energy multiplication factors have been obtained as 1.2727 and 1.3209, respectively, among 7.5 % and 90 % 6Li enrichment ratios. The thermal power, its ratio, and the hydrogen production amount have been calculated depends on the energy multiplication factor for various 6Li enrichment ratios and each considered cycle. The results showed that hydrogen production quantity increases with a rising 6Li enrichment ratio for both cycles. The highest produced hydrogen amount was obtained as 12.74 kg/s via the Cu-Cl cycle, whereas hydrogen production with the Co-Cl cycle has been computed as 6.79 kg/s in the cases of 10 % ThC fuel ratio and 90 % 6Li enrichment ratio.

Core neutronic design of small modular molten salt reactor for submarine propulsion

Molten salt reactor (MSR) shows great promise as a Generation IV nuclear reactor concept with high thermal-electric conversion efficiency, inherent safety features, and online reprocessing capability. During operation, the molten salt reactor MSR does not require a long refueling shutdown owing to online fuel reprocessing, making it suitable for watercraft propulsion. This paper presents the neutronic design of the 250 MWt iso-breeder MSR concept for submarine propulsion. The reactor is fueled with LiF-BeF2-ThF4-233UF4 salt with a virtual one-and-a-half fluid configuration. The MSR core is moderated with beryllium oxide and equipped with a high-density graphite reflector to improve neutronic. All calculations on the reactor core design were performed using MCNP6.2 code with ENDF/B-VII.0 neutron cross-section library. As a part of the neutronic analysis, the calculated parameters were the effective multiplication factor (), temperature coefficient of reactivity (TCR), and breeding ratio (BR). The value of ± 1.005 was obtained using Th : U ratio of 98.42% : 1.58%. The TCR value was obtained at -2.56 pcm/K, while the BR value was 1.057. These initial values indicate that the marine propulsion MSR can achieve iso-breeding with inherent safety characteristics. By using MSR design, it is hoped that the submarines will be able to reach significantly longer operational range than its Diesel-powered counterpart..

Exploratory neutronic evaluation on the enhancement of tritium breeding and energy multiplication capability using (Th,U)O2 in a solid-state tritium breeding blanket of a fusion demonstration reactor

The achievable tritium breeding ratio (TBR) in Korean fusion demonstration reactor (K-DEMO) that utilizing the deuterium-tritium nuclear fusion reaction for energy production is the critical design parameter for the self-sufficiency of tritium. However, the calculation uncertainty of the achievable TBR makes it difficult to fully understand the tritium fuel cycle in the K-DEMO. Moreover, if the achievable TBR is less than the required TBR, the calculation uncertainty of the achievable TBR makes the design limitation on the self-sufficient tritium supply of the K-DEMO. Therefore, making the achievable TBR larger than the required TBR is the most important goal of the K-DEMO design for the self-sufficient tritium supply. To create an achievable TBR with a margin greater than the required TBR, the modified neutron-multiplying material that is different from the typical neutron-multiplying material (9Be) is tested in this study. The blending of the fissionable (232Th) and fissile (233U) material with 9Be is attempted to enhance the neutron multiplication reaction resulting in a gradual increase of the achievable TBR according to the increase of the atomic faction of fissile material. With the addition of fissile material, it is found that energy multiplication accompanied by fission reaction is dramatic. The possibility of the self-breeding of 233U from 232Th is also justified. However, the neutronic damages resulting in the frequent replacement of the components in the tritium breeding blanket and generation of the high-level waste also occurred as the negative impact accompanied by the blending of 232Th and 233U with 9Be.

Compact fusion blanket using plasma facing liquid Li-LiH walls and Pb pebbles

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 1019 n/cm2, and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH2), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements.

Lithium lead titanate (Li2PbxTi1-xO3, 0.1

Tritium breeder is an important functional material in d-T fusion reactor blanket serving the purpose of tritium re-production and energy conversion. Presently, Li2TiO3 ceramics is considered as the most promising solid tritium breeder for its good chemical stability, high mechanical strength and excellent tritium release performance. However, low lithium content and lithium evaporation at high temperature lead to a low tritium breeding ratio. In the paper the neutron multiplier element, Pb, is introduced into Li2TiO3 to obtain lithium lead titanate (Li2PbxTi1-xO3) as tritium breeder, which combined the advantages of Pb and Li2TiO3 ceramic. Li2PbxTi1-xO3 powders were synthesized via a sol-gel method, employing lithium acetate, tetrabutyl titanate, and lead acetate, with varying lead contents. The characterization results indicate that at a molar ratio of titanium to lead of 1:2, Li2TiO3-Li2PbO3 eutectic phase could be formed. The tritium breeding ratio (TBR) of Li2PbxTi1-xO3 materials were evaluated by Monte Carlo Code, which was notably influenced by the lead content, exhibiting a continuous increase with higher Pb content and neutron energy. The Li2PbxTi1-xO3 pebbles were prepared by freeze-drying route. The mechanical strength of Li2PbxTi1-xO3 pebbles could be affected significantly by the lead content. Compared to the conventional Li2TiO3 ceramics, the Li2Pb0.4Ti0.6O3 ceramics exhibits a higher crushing load of 25.33 N.

Neutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixture

In this study, a novel approach has been investigated by using a mixture of thorium and molten salts as a dual-purpose coolant and medium for the production of fissile fuel in a Fusion Fission Hybrid Reactor (FFHR) for the reference geometry of ITER. The study highlighted the broader benefits of thorium fuel cycling, safety features, and reduced radioactive minor actinides generation. The use of a thorium-melted salt coolant for fissile fuel production in a fusion-fission hybrid reactor represented a promising path towards efficient and sustainable nuclear energy, with potential benefits in terms of safety features and reduced generation of radioactive minor actinides. In this study, SS 316 LN-IG was selected as the first wall material for the reactor, and a molten salt fuel mixture of LiF–ThF4 was used as the coolant, taking into account the eutectic points of the material, the nominal fusion power in the FFHR for the Tokamak design concept is considered to be 500 MW. The nuclear code MCNP6 was used with the nuclear data libraries ENDF/B-VIII and CLAW-IV for the neutron calculations. The time evolution of the isotopes in the reactor was calculated with the interface code MCNPAS. The study results are evaluated in terms of tritium breeding ratio, energy multiplication factor, radiation damage, fissile fuel production and fuel burn-up value.The 4-year operation history of total TBR value is calculated and always above 1.05 and increases with time.Th initially decreased from 631.3 tonnes to 587.2 tonnes, while 233U production during this period was 9.1 tonnes. According to these results, the first wall replacement period was calculated as 3.9 years.

Proof-of-principle of parametric stellarator neutronics modeling using Serpent2

This contribution presents neutron transport studies for the 5-period helical-axis advanced stellarator stellarator using the Serpent2 code. These studies utilize a parametric geometry model, enabling scans in neutronics modeling by varying the thickness of the reactor layers. For example, the tritium breeding ratio (TBR) can be determined by exploring various blanket material options and thicknesses to identify the threshold configuration that meets the TBR design criterion of 1.15. We found out that with the helium-cooled pebble ped candidate option, the TBR criterion is met with a breeding zone thickness of 26 cm, while with the dual-coolant lithium lead the threshold is exceeded at a thickness of 46 cm. Furthermore, the geometry includes non-planar field coils, allowing to study the fast neutron flux in these superconducting coils with a technological limit of 1×1091/cm2s . It is shown that the neutron fast flux is not constant at the coils, necessitating a neutron transport simulation to determine the distribution of the fast-flux at the coils. We show that the peak fast flux can be more than a factor of 2 higher than the average flux, and that the peak flux location rotates helically.

Utilization of Magnesium Chloride (Mg-Cl) cycle for Hydrogen Production of SOMBRERO Fusion Reactor

The aim of this study is to investigate the hydrogen production potential of the Mg-Cl (magnesium chloride) cycle in the SOMBRERO fusion reactor. Three different percentages of UO2 nuclear fuel, namely 2%, 6% and 10%, have been used while keeping the fuel zone thickness constant. The performance of the SOMBRERO fusion reactor has been considered statically. The neutronic calculations have performed with Monte Carlo neutron transport code. Firstly, it has been determined that the tritium breeding ratio (TBR) and energy multiplication factor (M) values depends on neutronically. Secondly, the amount of hydrogen production with by using energy multiplication factor (M) have been performed. The highest hydrogen production has been obtained as 8,12687 kg/s for 10% UO2 fuel ratio.

Performance evaluation of the Fe–Cl and Mg–Cl cycle for hydrogen production of the minor actinide fuelled PACER fusion blanket

Thermochemical processes used to produce clean hydrogen (H2) on a large scale by utilizing renewable energy and nuclear power sources are considered one of the most environmentally friendly strategies. In this study, the neutronic performance and hydrogen production potential of the PACER fusion blanket have been investigated during the operation period. Firstly, neutronic calculations have been performed to produce an electrical energy of 1000 MW from fusion explosions of 8.36 TJ to be repeated every 40 min with Monte Carlo Calculation Method (MCNP). Minor actinide flouride (MA) fuel out of the nuclear waste of LWRs with volume fractions of 2% have been mixed homogenously in the Flibe coolant. Tritium breeding ratio (TBR) at startup is calculated as 1.27. TBR and energy multiplication factor (M) decrease gradually. TBR >1.05 can be kept over 7 years for a self-sustained reactor. The M due to higher fissionable fuel content in the molten salt computes as 3.3 at initial. The burn up exceeds ∼800 GWd/tM at the end of operation period. Secondly, the potential of hydrogen production for four step iron chloride (Fe–Cl) and three step magnesium chloride (Mg–Cl) thermochemical cycles of the hydrogen unit PACER fusion blanket. The ratio of thermal power (1 − ψ), total thermal power (Phpf) and rate of mass flow of hydrogen (m˙H2) depends on M for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) thermochemical cycle have been presented. The ratio of thermal power (1 − ψ) for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) have been obtained as 0.6494, 0.131 and 0.095 whereas, total thermal power (Phpf) have been computed as 5298.8, 3608.43 and 3488.925 MW, respectively, at start up. The produced of hydrogen amount for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) have been calculated as ∼ 534.65; ∼1236.82 and ∼806.03 kg/year, respectively, at startup.

ParaStell: parametric modeling and neutronics support for stellarator fusion power plants

The three-dimensional variation inherent to stellarator geometries and fusion sources motivates three-dimensional modeling to obtain accurate results from computational modeling in support of design and analysis of first wall, blanket, and shield (FWBS) systems. Manually constructing stellarator fusion power plant geometries in computer-aided design (CAD) and defining the corresponding fusion source can be cumbersome and challenging. The open-source parametric modeling toolset ParaStell has been developed to automate construction of such geometries in low-fidelity. Low-fidelity modeling is useful during the conceptual phase of engineering design as a means of rapidly exploring the design space of a given device. The modeling capability of ParaStell includes in-vessel components and magnets, for any given stellarator configuration, using a parametric definition and plasma equilibrium data. Furthermore, the toolset automates the generation of detailed, tetrahedral neutron source definitions and DAGMC geometries for use in neutronics modeling. ParaStell assists rapid design iteration, parametric study, and design optimization of stellarator fusion cores. As a demonstration of the design iteration capability, the effect of the three-dimensional parameter space on tritium breeding and magnet shielding is investigated, using the WISTELL-D configuration as a design basis. Blanket and shield thicknesses are varied in three dimensions, using the space available between the plasma edge and magnet coils as a constraint. The corresponding effects on tritium breeding ratio and magnet heating are tallied using the open-source Monte Carlo particle transport code OpenMC. The inclusion of additional and higher-fidelity modeling capabilities is planned for ParaStell’s future, as well as its implementation in machine-driven optimization.

Tritium breeding ratio optimization in simple multi-layer blanket with genetic algorithm

Investigation on tritium production via fast neutrons generated by magnetically confined plasma in a tokamak device is conducted for optimization of tritium breeding ratio (TBR). The blanket is configured as a 1-dimensional multiple layer of materials for the wall, moderation, reflection, and tritium productions, and genetic algorithm is adopted to select the optimal material on each order and location. The configuration selected in the process is evaluated in aspect of tritium production to incoming neutron ratio. To construct the algorithm, the ratio of tritium production is parametrized by neutron energies from 0 to 14 MeV for the materials with Geant4 Monte Carlo simulation toolkit, and the simulated tritium in the algorithm are evaluated for the selection of parent for the next generation. Result configuration from the algorithm is put back to the Geant4 simulation for verification, and TBR is evaluated with blanket designs in other facilities.

Integral analysis of the effect of material dimension and composition on tokamak neutronics *

The neutronics performance of a tokamak has been identified as an important factor in designing a fusion power plant. The design of the tokamak should not only meet operational parameters such as sufficient tritium breeding, but also safety parameters such as low structural material activation. This paper investigates the impacts of the neutronics metrics for the ARC-class tokamak, a compact tokamak with an immersion blanket, by perturbing the first five layers of structural material—first wall, inner vacuum vessel, coolant salt channel, neutron multiplier, and outer vacuum vessel. The goal of this work is to provide insight into shaping and scaling the flux on each layer to obtain optimized operational and safety metrics through quantification of the responses from each perturbation. Results show that increased first wall thickness can increase the tritium breeding ratio (TBR) in specific configurations with high 6Li enrichments and that vacuum vessels decrease TBR for low-6Li enrichment configurations. It was also found that the neutron multiplier can either increase or decrease TBR depending on the configuration. The response of metrics to the change in layer thickness and enrichment also varies depending on the vacuum vessel material. The integral impacts of 6Li enrichment, layer thicknesses, and vacuum vessel material choice are investigated and presented in this paper.