Tritium Breeding Requirements Research Articles

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.

Modeling and analysis of the tritium fuel cycle for ARC- and STEP-class D-T fusion power plants

The limited tritium resources available for the first fusion power plants (FPPs) make fuel self-sufficiency and tritium inventory minimization leading issues in FPP design. This work builds on the model proposed by Abdou et al (2020 Nucl. Fusion 61 013001), which analyzed the fuel cycle (FC) of a DEMOnstration nuclear FPP-class FPP with a time-dependent system-level model. Here, we use a modified version of their model to analyze the FC of an Affordable, Robust, Compact (ARC)-class tokamak and two versions of a Spherical Tokamak for Energy Production (STEP)-class tokamak. The ARC-class tokamak breeds tritium in a 2LiF + BeF2 liquid immersion blanket, while the STEP-class tokamak breeds tritium utilizing either a liquid-lithium blanket design or an encapsulated breeding blanket. A time-dependent system-level model is developed in Matlab Simulink® to simulate the evolution of tritium flows and tritium inventories in the FC. The main goals of this work are to assess tritium self-sufficiency of the ARC- and STEP-class designs and to determine quantitative design requirements that can be used to analyze the adequacy of a proposed FC system. These design requirements are aimed at achieving a low tritium inventory doubling time () and a low start-up inventory () while keeping the required tritium breeding ratio (TBR) as low as possible. We also consider how improvements in FC technology and POs affect TBR and . The model results show that TBR for ARC- and STEP-class FPPs should be achievable if the tritium burn efficiency (TBE) reaches 0.5%–1% (TBR1.2). This assumes significant, but attainable, improvements over current abilities. However, the model results indicate that an FPP must achieve ambitious performance targets, including FPP availability 70%, tritium processing time 4 h, and the implementation of direct internal recycling (DIR). If future research yields major improvements to achievable TBE, it may be possible to achieve tritium self-sufficiency while operating at lower availability and without implementing DIR.

DEMO tritium fuel cycle: performance, parameter explorations, and design space constraints

One of the overarching goals of a DEMO-class device is to demonstrate tritium self-sufficiency in a fusion power plant for the first time. A future power reactor will necessarily require a start-up inventory of tritium, mTstart, before commencing fully fledged D-T operations for electricity production. In Europe, it is also presently considered necessary for DEMO to provide a tritium fuel start-up inventory for a subsequent prototype fusion power plant at a certain doubling time, td. At present, there is no model capable of estimating mTstart or td for the EU-DEMO, which features a Direct Internal Recycling (DIR) loop in its fuel cycle, and is characterised by low load factors (∼0.2-0.3). This paper introduces a simplified dynamic tritium fuel cycle model capable estimating mTstart and td, which has been specifically designed to take into account the effects of low reactor load factors and irregular operation. Results with and without DIR are presented. The fuel cycle design space is explored, and the sensitivity of the performance (in terms of mTstart and td) to variations in key parameters and parameter combinations is analysed. Minimum recommended values are suggested for the required tritium breeding ratio (Λr≥1.05), load factor (Aglob≥0.2), and DIR separation factor (fDIR≥0.6), for the assumptions made herein, based on the response of the fuel cycle performance in the explored design space.

Evaluation of the radiation damage parameters of ODS steel alloys in the first wall of deuterium-tritium fusion-fission (hybrid) reactors

One of the most important parameters in the design of the fusion-fission hybrid reactor is the selection of the first wall material. Because the oxide dispersion-strengthened (ODS) steel alloys have high temperature oxidation, high radiation resistance, good hardness, and corrosion resistance properties, they are thought first wall candidate materials for fusion and fission applications. The objective of this paper is to determine the best radiation damage parameters of various experimental and commercial ODS steels (namely, 12Y1, 12YWT, 1DS, IDK, Eurofer97, MA956, MA957, and PM 2000). Neutron spectrum and average neutron energy throughout blanket, displacement per atom, hydrogen and helium production, nuclear heating, and tritium breeding ratio were calculated by using Monte Carlo methods with Monte Carlo neutron-photon transport code and nuclear libraries named as ENDF/B-VI and CLAW-IV. It is assumed that calculated reactor has been operated full power during a year and neutron wall load is 2.25 MW/m2 (1014 n/s). All investigated first wall materials should be replaced between 3.5 and 4 years. All investigated materials provide minimum required tritium breeding ratio value, and when considering all the calculations performed in this work, 1DS ODS steel is the most suitable first wall materials with respect to other investigated ODS steels.

Effect of engineering constraints on charged particle wall heat loads in DEMO

The presently predicted total heating power of the demonstration fusion reactor DEMO is ∼3 times the ITER value, while the major radius is only 1.5 times larger (R. Wenninger, et al., The DEMO Wall Load Challenge, Nucl. Fusion 57 (2017) 046002 (11pp)) [1]. The current DEMO technological wall heat load removal capability is limited to ∼1MW/m2, due to structural material limitations and the tritium breeding requirements, while the ITER first wall (FW) is designed for values up to 4.7MW/m2. This paper focuses on the evaluation of the effect of the engineering constraints on the required limitation of charged particle heat load.First, a 2D field-mapping tool is used together with a simple model to take into account the peaking factors present on an engineering 3D wall design. A sensitivity analysis is performed on a set of realistic FW 3D features considering a preliminary estimate of misalignments. The impact on the heat flux to the wall due to different machine geometries, plasma shapes variation, stationary plasma and plasma transients, is presented. An automatic procedure to define the 2D poloidal contour of the FW for minimized particle loads is presented. A subset of the resulting engineering wall design is analyzed using the 3D field line tracing code PFCflux.

Tritium fuel cycle modeling and tritium breeding analysis for CFETR

Attaining tritium self-sufficiency is a critical goal for fusion reactor operated on the D–T fuel cycle. The tritium fuel cycle models were developed to describe the characteristic parameters of the various elements of the tritium cycle as a tool for evaluating the tritium breeding requirements. In this paper, a modified tritium fuel cycle model with more detailed subsystems and an updated mean residence time calculation method was developed based on ITER tritium model. The tritium inventory in fueling system and in plasma, supposed to be important for part of the initial startup tritium inventory, was considered in the updated mean residence time method. Based on the model, the tritium fuel cycle analysis of CFETR (Chinese Fusion Engineering Testing Reactor) was carried out. The most important two parameters, the minimum initial startup tritium inventory (Im) and the minimum tritium breeding ratio (TBRreq) were calculated. The tritium inventories in steady state and tritium release of subsystems were obtained.

Neutronic analyses of design issues affecting the tritium breeding performance in different DEMO blanket concepts

Neutronic analyses were performed to assess systematically the tritium breeding ratio (TBR) variations in the DEMO for the different blanket concepts HCPB, HCLL, WCLL and DCLL DEMOs due to modifications of the blanket configurations. A dedicated automated procedure was developed to fill the breeding modules in the common generic model in correspondence to the different concepts. The TBR calculations were carried out using the MCNP5 Monte Carlo code. The following parameters affecting the global TBR were investigated: TBR poloidal distribution, radial breeder zone depth, 6Li enrichment, steel content in the breeder modules, poloidal segmentation of the breeder blanket volume, size of gaps between blankets, thickness of the first wall and of the tungsten armour. Based on the results a set of practical guidelines was prepared for the designers developing the individual breeding blanket concepts with the goal to achieve the required tritium breeding performance in DEMO.

Investigation of tritium breeding ratio using different coolant material in a fusion–fission hybrid reactor

The aim of this study is to investigate tritium breeding ratio (TBR) and first wall material damage parameters such as displacement per atom (DPA), gas production of various coolants with Li2C2 tritium breeding material using the three different data libraries (ENDF/B.V, ENDF/B.VI and CLAW IV). Hence, FLiNaK, enriched (90% 6Li) natural lithium, lithium-lead eutectic alloy (Li17Pb83) and FLiBe (Li2BeF4) were used as a coolant (moderator) material for the calculation of the tritium breeding ratio and gas production parameters in a hybrid reactor. Neutronic calculations were performed by Monte Carlo Neutron-Particle Transport code MCNP5 version 1.40 under a neutron wall loud of 2.25 MW/m2 (1014 n/s) by full reactor power for one year. All examined coolant materials were achieved minimum required tritium breeding ratio (TBR > 1.05) and based on radiation damage limits. When the enriched (90%) lithium coolant with lithium carbide breeder used in fusion–fission (hybrid) reactor, vanadium alloy first wall will be replaced every 3.6 and 6.7 years for DPA and Helium production, respectively.

Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR)≥1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

ARIES-ACT2 DCLL Power Core Design and Engineering

ARIES-ACT2 is a conventional tokamak power plant conceptual design that uses a dual-coolant lead-lithium (DCLL) blanket concept with a RAFS (reduced-activation ferritic steel) first-wall (FW) and blanket structure. The design concept is the first fully integrated study of the DCLL blanket in a tokamak power plant. The major engineering efforts were to develop a credible configuration that can meet aggressive maintenance goals and achieve high availability and maintainability; to design a DCLL blanket that can meet tritium breeding requirements with reasonable helium and Pb-17Li cooling schemes to remove the surface and volumetric thermal power in the blanket while keeping the helium pressure drop, magnetohydrodynamic (MHD) pressure drop, and total pumping power low, and material temperatures and stresses at an acceptable level; to design manifolding and access pipes to connect/disconnect the inboard and outboard blanket sectors to the ring headers located underneath the reactor without affecting maintenance operations and creating major MHD effects when feeding all the Pb-17Li/He mass flow. Detailed three-dimensional finite element analysis of the DCLL blankets together with design iterations have been performed to finalize and optimize the major design parameters of the FW and blanket structure. The helium-cooled W plate-type divertor concept was adopted and integrated into the ACT2 DCLL power core to accommodate the peak surface heat flux of ∼10 MW/m2 predicted by edge plasma physics.

Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

LIFE Tritium Processing: A Sustainable Solution for Closing the Fusion Fuel Cycle

The Laser Inertial Fusion Energy (LIFE) power plant is being designed to deliver a transformative source of safe, secure, sustainable electricity, in a time scale that is consistent with the global energy market needs. The LIFE market entry plant will demonstrate the feasibility of a closed fusion fuel cycle, including tritium breeding, extraction, processing, re-fueling, accountability and safety, in a steady-state power-producing device. While many fusion plant designs require large quantities of tritium for startup and operations, a range of design choices made for the LIFE fuel cycle act to reduce the inprocess tritium inventory. The high fractional burn-up (~30%) in an Inertial Fusion Energy (IFE) capsule relaxes the tritium breeding requirements, while the use of only milligram quantities of fuel per shot and choice of a pure lithium heat transfer fluid substantially reduce the amount of material entrained in the facility. Additionally, the high solubility of tritium in the lithium breeder is expected to mitigate the need for development of permeation barriers in the engine systems, normally required to control routine releases within the allowable regulatory limits.The present paper offers an overview of the design of the LIFE fuel cycle, including a summary of the technology development plan consistent with the delivery schedule of the LIFE market entry plant.

Molten salt self-cooled solid first wall and blanket design based on advanced ferritic steel

Abstract As an element in the U.S. Advanced Power Extraction (APEX) program, the solid first wall and blanket design team assessed innovative design configurations with the use of advanced nano-composite ferritic steel (AFS) as the structural material and FLiBe as the tritium breeder and coolant. The goal for the assessment is to search for designs that can have high volumetric power density and surface heat-flux handling capability, with assurance of fuel self-sufficiency, high thermal efficiency and passive safety for a tokamak power reactor. We selected the re-circulating flow configuration as our reference design. Based on the recommended material properties of AFS we found that the reference design can handle a maximum surface heat flux of 1 MW/m2, and a maximum neutron wall loading of 5.4 MW/m2, with a gross thermal efficiency of 47%, while meeting all the tritium breeding, structural design and passive safety requirements. This paper will cover the results of the following areas of assessment: material design properties, FW/blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal-hydraulics analysis including MHD effects, structural analysis; molten salt and helium closed cycle power conversion system; and safety and waste disposal of the re-circulating coolant first wall and blanket design.

APEX advanced ferritic steel, Flibe self-cooled first wall and blanket design

As an element in the US Advanced Power Extraction (APEX) program, we evaluated the design option of using advanced nanocomposite ferritic steel (AFS) as the structural material and Flibe as the tritium breeder and coolant. We selected the recirculating flow configuration as our reference design. Based on the material properties of AFS, we found that the reference design can handle a maximum surface heat flux of 1 MW/m 2, and a maximum neutron wall loading of 5.4 MW/m 2, with a gross thermal efficiency of 47%, while meeting all the tritium breeding and structural design requirements. This paper covers the results of the following areas of evaluation: materials selection, first wall and blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal hydraulics analysis including MHD effects, structural analysis, molten salt and helium closed cycle power conversion system, and safety and waste disposal of the recirculating coolant design.

A New Approach for Assessing the Required Tritium Breeding Ratio and Startup Inventory in Future Fusion Reactors

Accurately estimating the required tritium breeding ratio (TBR) λr in fusion reactor systems is necessary to guide fusion research and development and to assess the feasibility of fusion reactors as a self-sufficient energy source. This is especially true when one considers the limits imposed by the present-day breeding performance of breeder blanket candidates. Studies of this subject have been performed in the past, with particular emphasis on developing appropriate dynamic simulations of the fuel cycle. In the last few years, development of new dynamic and integrated fusion fuel cycle tritium computer codes has moved away from general residence-time models and instead incorporated more comprehensive and realistic models. Furthermore, detailed and rigorous computer codes that model the dynamic retention behavior of individual components inside the fuel cycle, in particular the torus plasma-facing components in a tokamak, have been vastly improved with uncertainties identified. A more efficient and intuitive methodology for tritium self-sufficiency analyses is developed based on an analytical scheme that makes use of different types of tritium inventories inside the fuel cycle as calculated from detailed numerical simulations. Short-term and long-term tritium inventories are differentiated as well as tritium lost through waste material. Also, the tritium fuel cycle is split into a number of independent tritium migration paths to aid in the development of an integrated tritium balance for which λr or other parameters of interest can be solved analytically. Tritium startup requirements are also examined. An important side benefit derived from using the aforementioned methodology is that the uncertainty in λr for a given reactor design can easily be calculated from uncertainty ranges characterizing a number of relevant reactor operation and fuel cycle parameters. Maximum tritium inventory limits were considered from safety and operational standpoints. A wide range of parametric studies were conducted with various scenarios to forecast changes in λr when the reactor design is modified. For example, it was determined that with most current estimates of the achievable TBR λa, ranging from 1.04 to 1.07, a small design window for both the fuel fractional burnup and the downtime of tritium reprocessing components severely limits any proposals for a reactor operating scenario that will be valid for a reasonably paced fusion growth rate.

Materials integration issues for high performance fusion power systems

One of the primary requirements for the development of fusion as an energy source is the qualification of materials for the first-wall/blanket system that will provide high performance and exhibit favourable safety and environmental features. Both economic competitiveness and the environmental attractiveness of fusion will be strongly influenced by the materials constraints. A key aspect is the development of a compatible combination of materials for the various functions of structure, tritium breeding, coolant, neutron multiplication and other special requirements for a specific system. This paper presents an overview of key materials integration issues for high performance fusion power systems. Issues such as: chemical compatibility of structure and coolant, hydrogen/tritium interactions with the plasma facing/structure/breeder materials, thermomechanical constraints associated with coolant/structure, thermal-hydraulic requirements, and safety/environmental considerations from a systems viewpoint are presented. The major materials interactions for leading blanket concepts are discussed.

Fuel Source Isotopic Tailoring and Its Impact on International Thermonuclear Experimental Reactor Design, Operation, and Safety

Isotopic tailoring of the deuterium and tritium density profiles in fusion reactors can lead to reduced tritium inventory in plasma facing components and, therefore, improved safety considerations. The isotopic tailoring concept consists of utilizing a tritium-rich pellet source for core fueling and a deuterium-rich gas source for edge fueling. Because of the improved particle confinement associated with the deeper tritium core fueling component, comparable core densities of deuterium and tritium can be maintained even when the edge deuterium fuel source is much larger than the core tritium fuel source. The fuel composition of the edge and scrape-off plasmas as well as the isotope fractions in plasma facing components reflect the total throughput of all makeup fuel and are therefore deuterium-rich. This innovative fueling concept results in about a factor of two reduction in tritium inventory of the plasma facing components. The higher tritium burn fraction allows a significant reduction in tritium gas flows into and out of the vacuum vessel and, for fusion reactors, implies lower required tritium breeding ratios. 10 refs., 3 figs., 2 tabs.

Three-dimensional analyses of candidate blanket designs for the compact tokamak reactor concept

Three-dimensional nucleonic analyses have been performed for five candidate blanket systems proposed for a compact tokamak fusion reactor. These analyses were carried out using the Monte Carlo code MCNP and the nuclear data ENDF/B-IV. It has been found that the ternary ceramic breeder LiAlO 2 does not meet the required tritium breeding goal (TBR=1.1) even when the best nonfissionable neutron multiplier and different design configurations were utilized. Liquid metal breeders have better nucleonic performance than the solid breeders. Liquid lithium is superior to Li 17Pb 83 in terms of nucleonic performance especially when no breeding in the inboard is utilized and when peneteration systems are considered. Complete elimination of inboard breeding and accounting for penetration systems resulted in breeding losses of 23.0%, 25.3%, 23.5% and 32.9% in blankets of Li 2O/H 2O/SS316L, Li 2O/He/SS316L, Li/Li/V15Cr5Ti and Li 17Pb 83/Li 17Pb 83/HT-9 respectively. A careful choice of limiter structure and coolant leads to less breeding loss. The effect of the RF-system on the TBR was small for all the blanket systems considered. When Li 17Pb 83 is used as a static breeder to minimize its circulation, and thus making MHD losses negligible, a separate coolant such as water, helium or sodium should be used. It has been found that when high 6Li-enrichment is used the effect of the coolant choice on the TBR and energy multiplication factor, EMF, is negligible. The TBR of the Li 2O blanket cooled with water and without inboard breeding can be enhanced to 1.28 which is large enough to compensate for the losses associated with penetration systems. The atomic displacement rate in the first wall of the Li 17Pb 83 blanket system is the largest followed by that of the Li-blanket, with rates in the solid breeder blankets about 10% lower than that of Li 17Pb 83. The helium production rate in the first wall of a LiAlO 2 blanket system is the largest and it is high in the structure of the front breeder zones of solid breeder blankets. The hydrogen production rate in the first wall and front breeder structures is also high for solid breeder systems.

Solid breeder blanket design and tritium breeding

Thermonuclear D–T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements—to breed one tritium per fusion neutron—are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li 2O, LiAlO 2, Li 2ZrO 3, Li 4SiO 4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket.

Determination of the Required Tritium Breeding Ratio for the Next Fusion Experimental Reactor

The next fusion experimental reactor such as ITER requires tritium breeding because of the high tritium cost and its limited availability from non-fusion sources, in addition to demonstrating breeding capability of commercial D-T reactors. A tritium fuel cycle model was developed to compute the required tritium breeding ratio (TBR) by taking into account rector down-time. The results show that TBR should be unity to achieve 3 MW * Year/m/sup 2/ of neutron fluence in 10 years for a steady-state reactor with 600 MW fusion power and 25% system availability provided 5 kg of initial tritium supply. If the external tritium supply is increased to 20 kg, the required TBR is 0.9. The estimated TBR is very sensitive to the variation of the tritium burn-up fraction in plasma and the tritium residence time in the tritium processing system. For example, decreasing the burn-up fraction from 5% to 1% leads to a 25% increase in the required TBR. Thus these parameters must be carefully examined in future work.