Nuclear Fusion Energy Research Articles

Current status of research on magnetic confinement fusion and superconducting tokamak devices

Superconducting tokamaks have garnered significant research and interest in the quest for harnessing nuclear fusion energy. They are considered one of the most promising devices for achieving this goal. After more than half a century of dedicated efforts, a group of international researchers has successfully constructed and operated a batch of tokamak devices, resulting in significant achievements in magnetic confinement fusion research. The purpose of this paper is to provide a comprehensive review of the development of tokamak devices and magnetic confinement fusion. Firstly, the principles, main research directions and prospects of magnetic confinement fusion are introduced. It then reviews the construction and operation of major devices in the international context, including the International Thermonuclear Experimental Reactor (ITER), and describes the purpose and main contributions of their construction, and analyses the latest research results and development trends in this field from the perspective of experimental research progress. The current challenges and future directions of superconducting tokamak devices are also discussed. The review will provide a comprehensive and systematic introduction and reference for researchers and engineers in related fields.

Li6BaLa2Ta2O12 Solid-State Probe for Studying Li Activity in Molten Sn-Li Alloys

Metals or alloys that are liquid at or near room temperature are of interest of the scientific community due to the versatility of their applications. Considering the nuclear fusion energy applications, Sn-Li alloys appear as a great candidate for plasma facing materials. However, plasma interactions can induce alterations on the lithium concentration and, as a consequence, modify the properties of the LM. Therefore, in order to prevent these situations, thermodynamic data such as activity coefficients and analytical devices capable of monitoring the lithium content in these alloys are of great interest. In this work, a lithium sensor based on the Li6BaLa2Ta2O12 solid-state electrolyte was used to determine lithium activity correlations for Sn-Li alloys at 400 and 500 °C. It was observed that the activity coefficient was constant for lithium concentrations lower than 8.5 at% with values, at the infinite dilution condition, of 8.4·10−5 at 400 °C and 2.7·10−4 at 500 °C, respectively. Above this concentration (from 8.5 to 16.5 at% Li), the activity coefficient varied with lithium concentration. In this case, correlations between lithium activity and the lithium atomic fraction were obtained.

Proton-beam driver transport in the fast ignition of proton–boron-11 fuel plasma

The invention of Chirped Pulse Amplification lasers along with the development of fast ignition methods have made fuels such as proton–boron-11 seem promising nuclear fusion energy sources because of their lack of high-energy neutrons and abundance of constituent isotopes. This paper investigates the transport of charged particles in proton–boron-11 fuel. In this study, the confinement parameter has been estimated as [Formula: see text] g/cm2 with the help of a proton driver in the fast ignition method for proton–boron-11 fuel pellets, with a ratio of boron-11 to proton [Formula: see text]. Subsequently, using the Li–Petrasso stopping-power model, the energy deposition coefficients of the accelerated proton-beam driver were calculated by the Target Normal Sheath Acceleration method with 1 MeV energy. The obtained results show that at a temperature of about 65 keV, the proton-beam driver with 1 MeV energy deposits about 85% of its energy in the hot spot area of 2.5 [Formula: see text]m depth from the surface of the fuel pellet to ignite the proton–boron-11 fuel pellet.

Layer-structured Cr/CrxN coating via electroplating-based nitridation achieving high deuterium resistance as the hydrogen permeation barrier

Hydrogen isotope permeation through structural materials is a key issue for developing nuclear fusion energy, which will cause fuel loss and radioactive pollution. Developing ceramic coatings with high thermal shock and hydrogen resistance is an effective strategy to solve this issue. In this work, a layer-structured Cr/CrxN coating was successfully fabricated by a facile electroplating-based nitridation technique, which is easy, facile, and applicable to coating complex-shaped substrates. The Cr/CrxN coating, composed of a bottom Fe/Cr interdiffusion zone, a middle Cr layer, and a top CrxN layer, exhibits high bonding strength, high anti-thermal-shock ability, and high deuterium permeation resistance. Its bonding strength achieves 43.6 MPa. The Cr/CrxN coating remains intact even after suffering 300 thermal shock cycles under a 600 °C-water condition. Through optimizing the nitridation temperature, the Cr/CrxN coating achieves a deuterium permeation reduction factor (PRF) as high as 3599 at 500 °C. Considering its scalable fabrication technique and considerable properties, the developed Cr/CrxN coating may serve as a novel high-performance hydrogen permeation barrier in various fields.

Recent Progress in Research of Solid Tritium Breeder Materials Li2TiO3: A Review

During the past decades, fusion reactor fuels such as deuterium and tritium have been extensively investigated due to increasing interest in nuclear fusion energy. Tritium, which is scarce in nature, needs to be fabricated by tritium breeder materials. Among the commonly investigated tritium breeder materials, lithium titanate (Li2TiO3) is recognized as one of the most promising solid tritium breeder materials because of its considerable lithium (Li) atomic density, low activation, excellent chemical stability, and low-temperature tritium release performance. This paper aims to provide a systematic review of the current progress in Li2TiO3 preparation methods as well as the high Li density, tritium release performance, irradiation behavior, and modification technologies of Li2TiO3 pebbles. Li2TiO3 can be synthesized by strategies such as solid-state, sol–gel, hydrothermal, solution combustion synthesis, and co-precipitation methods. Among them, the hydrothermal method is promising due to its simplicity and low cost. Many researchers have begun to focus on composite ceramic pebbles to further improve tritium breeder performance. This will provide a new direction for the future development of Li2TiO3 pebbles. The present review concludes with a summary of the preparation methods currently under development and offers an outlook of future opportunities, which will inspire more in-depth investigation and promote the practical application of Li2TiO3 in this field.

Modeling Isotope Separation in Electrochemical Lithium Deposition

Naturally occurring Li consists of two stable isotopes, 6Li with an abundance of about 7.5%, and 7Li making up the remainder with 92.5%. The development of a 6Li enrichment technique, in terms of technical reliability and environmental safety to reach 6Li future requirements, represents a key step in the roadmap for nuclear fusion energy supply worldwide. This paper uses finite element analysis-based models to simulate electrochemical Li isotope separation, which is an attractive method in terms of simplicity, safety, and scalability. In the model, we quantitatively analyze how different electrochemical factors including thermodynamics, charge-transfer kinetics, and diffusivities affect the separation process (separation factor), together with cell parameters, such as cell length and current density. The maximum separation factor of 1.128 could be obtained with the cell under the optimal thermodynamic, kinetic, and diffusive conditions, which is among the highest separation factors ever reported. These results will assist in designing the actual isotope separation setup with large separation factor and appropriate timing for sample collection.

Lithium orthosilicate as nuclear fusion breeder material: Optimization of the drip casting production technology

The production of nuclear fusion energy, in the future DEMOnstration power plant (DEMO), is based on tritium self-sufficiency, achievable exploiting neutron-lithium reaction occurring in the breeding blanket modules. To fulfil the plant design conditions ceramic materials containing lithium are suggested. A lot of research efforts focused on the past years on the investigation and selection of the most suitable breeding material.In this study, the lithium orthosilicate, in pebble form, produced at room temperature at the University of Pisa was investigated and characterized from a thermo-mechanical point of view. The pebbles were produced by a drip casting forming technique, starting from an aqueous suspension of a Li4SiO4 precursor obtained by hydrolytic sol–gel method, and dropping it into a calcium chloride/lithium acetate solution. A high temperature thermal treatment was applied to green pebbles to synthesize Li4SiO4 and densify the structure. Thermal analysis and static uniaxial compression tests, without radial constraints, were carried out on different types of pebbles produced varying the composition of the precursor suspension by means of cellulose fiber additions. Results show the influence of the cellulose content on morphological and physical properties of the obtained pebbles. Li4SiO4 pebbles with a density of 88 % and a crush load of about 30 N were obtained when the suspension was modified by adding 15 wt% of cellulose fibers to the sol–gel precursor powders.

Status of nuclear fusion energy science & technology for power production: progress & prospects

Energy release by nuclear fusion reactions was first proposed in the context of theories explaining the energy source of stellar objects. Subsequently, laboratory experiments in the 1930's demonstrated the validity of such reactions. From the 1940's onward, efforts began to pursue the controlled & uncontrolled release of nuclear fusion energy for both civilian & military applications. While the military aspect of this technology was proved in the early 1950's, the civilian part, which focused on inventing different techniques to fuse light elements nuclei to produce power, found little success. The main reason for this was undertaking expensive & complicated fusion experiments without fully resolving all the scientific issues regarding each concept. However, the decades – long delays in realizing fusion for power production caused some of these efforts to shift recently & noticeably to the commercial arena where a literal accountability of the progress of these projects is conducted. This status review paper gives a quick historical background on nuclear fusion science & technologies, explains briefly a few innovative nuclear fusion energy production methods with a focus on the rapidly growing commercial trend of nuclear fusion research and developments efforts.

Nuclear energy basics

In the first part of the paper, the basic features of nuclear fission and fusion energy are addressed. In the second part, an overview of energy consumption in the world is presented, which highlights the growing need for electricity from non-CO2 emitting sources, among which nuclear power is expected to play a role. The perspectives for the future of nuclear energy are then discussed, showing that it develops mostly in emerging countries and the small modular reactors are likely to have an increasing role. The third part is devoted to the question of nuclear waste management and to the non-radioactive environmental impact of nuclear energy in comparison with other sources of energy.

Code O-SUKI-N 3D: Upgraded direct-drive fuel target 3D implosion code in heavy ion inertial fusion

The Code O-SUKI-N 3D is an upgraded version of the 2D Code O-SUKI (Sato et al., 2019 [1]). Code O-SUKI-N 3D is an integrated 3-dimensional (3D) simulation program system for fuel implosion, ignition and burning of a direct-drive nuclear-fusion pellet in heavy ion beam (HIB) inertial confinement fusion (HIF). The Code O-SUKI-N 3D consists of the three programs of Lagrangian fluid implosion program, data conversion program, and Euler fluid implosion, ignition and burning program. The Code O-SUKI-N 3D can also couple with the HIB illumination and energy deposition program of OK3 (Ogoyski et al., 2010 [8]). The spherical target implosion 3D behavior is computed by the 3D Lagrangian fluid code until the time just before the void closure of the fuel implosion. After that, all the data by the Lagrangian implosion code are converted to the data for the 3D Eulerian code. In the 3D Euler code, the DT fuel compression at the stagnation, ignition and burning are computed. The Code O-SUKI-N 3D simulation system provides a capability to compute and to study the HIF target implosion dynamics. Program summaryProgram Title: O-SUKI-N 3DCPC Library link to program files:https://doi.org/10.17632/h8n474wvf5.2Licensing provisions: CC BY NC 3.0Programming language: C++Journal reference of previous version: Comput. Phys. Commun. 240, 83 (2019)Nature of problem: Nuclear fusion energy would be energy source for society. In this paper we focus on heavy ion beam (HIB) inertial confinement fusion (HIF). A spherical mm-radius deuterium (D) - tritium (T) fuel pellet is irradiated by HIBs to be compressed to about a thousand times of the solid density. The ion temperature of ∼5-10 KeV is required to ignite the DT fuel to release a sufficient fusion energy. The typical HIBs total input energy is several MJ, and the HIBs pulse length is about a few tens of ns. The O-SUKI-N 3D code system provides an integrated tool to simulate the HIF DT fuel pellet implosion, ignition and burning in 3 dimensions (3D). The O-SUKI-N 3D code system is an upgraded version of the Code O-SUKI (Sato et al., 2019 [1]) which is a 2D implosion simulation system in HIF. The DT fuel is compressed to the high density of about a thousand times of the solid density, and so the DT fuel spatial deformation may be serious at the DT fuel stagnation. Therefore, the O-SUKI and O-SUKI-N 3D systems employ a Lagrangian fluid code first to simulate the DT fuel implosion phase until just before the stagnation. Then all the simulation data from the Lagrangian code are converted to them for the Euler fluid code, in which the DT fuel ignition and burning are simulated.Solution method: In the two fluid codes (Lagrangian and Euler fluid codes) in the O-SUKI-N 3D system the three-temperature fluid model (Tahir et al., 1986 [10]) is employed to simulate the pellet dynamics in HIF.Additional comments including restrictions and unusual features: The Lagrange code is inherently weak against the spatial mesh deformation due to the implosion non-uniformity. When short-wavelength perturbations are imposed near the poles of the spherical target, the spatial meshes might crash and the computation run may stop.

Numerical Simulations of Sub-Atmospheric J-T Heat Exchanger for Superfluid Helium Cryogenic System

Superfluid helium has been used for cooling superconducting magnets in nuclear fusion and high energy physics. This paper presents a simplified final stage process of 1.8 K superfluid helium cryogenic system and its key component sub-atmospheric heat exchanger. Influence of variable thermal parameters of sub-atmospheric heat exchanger on the system performance is studied. The application of sub-atmospheric heat exchanger can increase refrigerating capacity by 61.34% under given conditions. The helically finned-tube heat exchanger (HFHE) is selected, and three-dimensional steady-state numerical simulations of flow and heat transfer performance have been carried out. Nu and f of tube side are 4.5-6 times higher and 95-97% lower than those of shell side, respectively. It is found that HFHE with low fin height and HFHE with filler (nylon rope) have better comprehensive heat transfer and flow performance, and PEC can reach 1.65 and 1.41 respectively under given conditions.

A Carbon-Free Energy Future on the Horizon

The UK is playing a leading role in the development of nuclear fusion energy and new technological advancements have been made. The potential energy source is seen as the answer to a carbon-free sustainable future

Electrochemical Response of Sr(Ce0.9Zr0.1)0.95Yb0.05O3-α high-Temperature Hydrogen Sensors

In future fusion reactors, tritium and deuterium will be used as a primary fuel. Currently, the most important research and development project about nuclear fusion energy is ITER (International Thermonuclear Experimental Reactor). To produce the fusion reaction, tritium will be generated inside the core using a breeding blanket. One of the suggested blankets is based on the use of molten lithium–lead eutectic alloy (Pb-15.7Li) [1]. This blanket will be flowing around the reactor core and while the fusion reaction occurs, energy is released and collected with an external heat exchanger system.Due to its short half-life (12.3 years), there are not natural tritium sources. Therefore, it must be obtained in situ and continuously in the breeding blanket. This hydrogen isotope can be produced by bombarding lithium with fast neutrons, which escape from the generated and confined plasma in the toroid-shaped vacuum chamber of a tokamak, to yield tritium and helium (6Li + n → 4He + 3H + 4.8MeV) [2]. Hence, dynamic tritium concentration measurement is of major interest for nuclear fusion reaction engineering and experimental proof of tritium self-sufficiency in liquid metal breeding systems. Consequently, it is necessary a measuring system that allows determining tritium concentration in high-temperature conditions.Electrochemical sensors based on solid-state proton conductor ceramics can be used for that purpose. These materials have attracted significant interest because of their chemical and physical durability, especially at elevated temperatures. These electrolytes are perovskite-type materials with electrical carriers, positive holes, excess electrons, oxide ion vacancies, and interstitial protons, which interact with oxide ions. Under this approach, new methods of synthesis are being studied in order to obtain ceramic materials of higher quality and purity able to offer better measurement signals.In the present work, the proton-conducting perovskite Sr(Ce0,9-Zr0,1)0,95Yb0,05O3-α was synthesized by two different methods. On one side, a conventional sol-gel method, using citric acid as a complexing agent, was employed [3]. On the other hand, the second method used was the glycine method, which is based on the combustion of glycine with the precursor's nitrates. Using the glycine method, it could be obtained a ceramic porous foam that improved the sintering process. Both synthesized materials were characterized using X-Ray Diffraction and SEM-EDS analysis.Finally, the obtained ceramic materials were properly sintered and shaped in order to construct electrochemical hydrogen sensors. Amperometric measurements were performed at 500°C and 600°C using different hydrogen concentrations and the response of the two sensors was compared.

The Digital Assembly and Maintenance Training Platform for ITER-Type Mock-Up in Virtual Reality Environment

The International Thermonuclear Experimental Reactor (ITER) is one of the largest international cooperative projects with many participants from different research groups in institutes and universities all over the world and is expected to achieve nuclear fusion energy output and to be operated in steady state with tritium self-sufficiency. The conventional design of the ITER-type facility always directly faces challenges of operation difficulties in the assembly and maintenance processes considering that there are collision interference and corresponding precision problems within the constrained space; consequently, it is not possible to provide for research purposes qualitative information for human intuition and quantitative contents with professional advice. Therefore, it is necessary to find a better solution for researchers and operators to get a highly efficient group-based work form without being blocked for reasons of geography. In this context, virtual reality technology has been introduced in the digital assembly and maintenance training platform of the ITER-type mock-up at the Institute of Modern Physics, Chinese Academy of Sciences, and it has many useful features, including real-time collision detection in assembly tasks with simplified trigger mesh structures by means of the ray-casting method, and three-dimensional visualization of the topological structure in the welding and brazing maintenance using the UV unwrapping and remapping methods considering the conversion process from digital values to gray-scale texture, which can meet the flexible and diverse design requirements and provide feasibility of training at the University of Chinese Academy of Sciences.

SMART_TC: an R&D Programme on uses of artificial intelligence techniques for tritium monitoring in complex ITER-like tritium plant systems

The realization of nuclear fusion energy is nowadays based on the concept of tritium breeding and the success of the ITER experiment. The latter relies today on a static monitoring approach to fulfill the emission limits imposed by the regulatory institutions. Artificial intelligence applications for fault diagnosis and process monitoring anticipate potential for the dynamic management of tritium in complex plant systems. This paper explores the dynamic tritium inventory management issue in complex systems, reviews the diverse artificial intelligence techniques and discusses the most promising approaches for ITER-like plant system match balance monitoring.

Evolution of Defects in CVD-W Irradiated by H/He Neutral Beam Using Positron Annihilation Spectroscopy

One of the key problems for the application of nuclear fusion energy is to select the suitable plasma facing materials (PFMs). Among the W-based materials, CVD-W exhibits some unique advantages. In order to estimate the performance of CVD-W under the fusion environment, the vacancy-type defects and their evolution are investigated by the Doppler-broadening slow positron beam analysis (DB-SPBA) combined with SEM (scanning electron microscope). There are two kinds of neutral beam irradiation, the pure H neutral beam and the H + 6 at.% He neutral beam irradiation, which are performed at the neutral beam facility GLADIS (IPP, Germany). The surface temperatures of CVD-W irradiated by H (H + 6 at.% He) are 850 and 1000 (700 and 800 °C). By comparing the samples under different conditions, the defect evolution of CVD-W is obtained. As for the pure H neutral beam irradiated samples, the DB-SPBA results demonstrate that the CVD-W sample at the surface temperature of 1000 °C, compared to the 850 °C sample, shows a decrease in S parameters, which is due to the reduction of vacancy-type defect concentration. The defect damage layer in 1000 °C sample is narrower than that of 850 °C sample and the defect type tends to be consistent in 1000 °C sample. The SEM results suggest that the surface damage of the 1000 °C sample was recovered to some extent. As for the H + 6 at.% He neutral beam irradiated samples, compared with the CVD-W sample at the surface temperature of 700 °C, the 800 °C sample shows an increased S parameters, which can be attributed to the volume increase of vacancy-type defect. The defect damage layer in the 800 °C sample is wider than that of the 700 °C sample. Both the H + 6 at.% He irradiated samples show complex defect types. The surface of the 800 °C sample exhibits more dense pinhole damage structures compared to that of the 700 °C sample.

The progress and current status of Tokamak: a systematic review

Nuclear fusion energy is an ideal energy source for the future due to the clean and efficient features, first proposed by Russian scientists in the 1950s. With the successful construction and operation of a batch of tokamak devices globally, a series of major achievements have been realized in magnetic confinement fusion (MF). On this basis, the recent progress and current status of tokamak devices are systematically reviewed from academic literatures and official websites. To better demonstrate, the principle of MF and the history of a tokamak are introduced firstly. Subsequently, developments of four typical state-of-art tokamak facilities (JT60, FTFR, JET, EAST) are discussed detailly. These results offer a guideline for tokamak device construction and MCF realization.

Mechanism of chip formation and surface-defects in orthogonal cutting of soft-brittle potassium dihydrogen phosphate crystals

Micromachining repair of surface defects on KH2PO4 (KDP) optics is an emerging technique in the construction of Inertial Confinement Fusion facilities for obtaining clean nuclear fusion energy. However, this method is yet facing considerable challenges owing to the soft-brittle nature of single-crystal KDP, hence it is necessary to understand its ductile-regime cutting mechanism to generate crack-free surfaces. This paper seeks to investigate the evolution of different cutting mechanism with the change of uncut chip thickness (UCT) in KDP orthogonal cutting processes. A transition of cutting modes from plastic cutting to shear-crack cutting and then fracture cutting with the rise of UCT has been revealed. To explain these cutting phenomena, a novel theoretical model was proposed by calculating the specific energy dissipation for crack/fracture propagations during cutting processes based on fracture mechanics. This analytical model was well validated by the analysis of cutting forces and machined surface quality. Nevertheless, three kinds of surface defects have been observed, i.e. micro pits, micro craters and edge chipping. These surface defects were caused by tearing and spalling of materials with elastic recovery, crack propagation along cleavage planes with ploughing effect, and the peeling away of large-size fracture, respectively. The presented results of great significance for promoting the application of micromachining processes in future engineering repair of KDP optics.

Towards Fusion Energy in the Industry 5.0 and Society 5.0 Context: Call for a Global Commission for Urgent Action on Fusion Energy

This article provides a high-level review of the geopolitical status quo of nuclear fusion energy and outlines a vision for the future against the backdrop of Industry 5.0 and Society 5.0. In 2020, the year of the 75th anniversary of Hiroshima and Nagasaki, ITER, the world’s largest fusion experiment, began its machine assembly, and next year, the DEMO (fusion power) stage of ITER planning is formally due to begin. With several countries now engaged in the design of DEMO-phase machines, we stress that a “burning plasma” self-sustaining fusion reaction event will be a critical juncture similar in status to the Trinity Test. In response, urging the need for US geopolitical engagement, we call for a “Global Commission for Urgent Action on Fusion Energy,” to be backed by the International Atomic Energy Agency and International Energy Agency (IEA), along the lines of the existing IEA “Global Commission for Urgent Action on Energy Efficiency.” We see the Commission serving a similar, this time international, role to the 1946 “Acheson-Lilienthal” report for fission and involving a global External Independent Review of the state-of-the art of fusion energy technologies. We suggest major “soft power” co-chairs from the Global West and North and from the Global South, thereby providing a geopolitical pivot which can unlock tens of billions of dollars in funding for multiple technology pathways, create a global regime to safeguard Intellectual Property, accelerate the transition from fossil fuels to a “Future Fusion Economy,” and prevent the geopolitical polarization of fusion energy.

AC loss and contact resistance of different CICC cable patterns: Experiments and numerical modeling

• The AC loss and interstrand contact resistance measured on three prototype Nb 3 Sn Cable-In-Conduit Conductors with significantly different cabling twist pitch patterns. • Simulations, done by using the JackPot model are in fair agreement, confirming the consistency of data and allowing reliable predictions with the JackPot model. • For CICC cables, a design petal wrap coverage of 70 % should be maintained for CS cables to reduce the AC loss. A critical threshold value of void fraction exists, which makes the presence of petal wraps ineffective in reducing coupling loss. • The JackPot simulations with derived corrections for inter-petal contact resistance representing 70 % petal wrap coverage show that the Twente cable design has the lowest coupling loss. • So far, the Twente cable design seems a suitable candidate for CFETR or DEMO conductors, made of strain-sensitive strands such as Nb 3 Sn or other materials, for minimization of coupling loss and strand indentation. For upcoming nuclear fusion energy reactors, like the China Fusion Engineering Test Reactor (CFETR) and EU-DEMO, the superconducting Cable-In-Conduit Conductors (CICC) are in the design phase, and the operating conditions like electromagnetic forces can be higher than in previous devices like ITER. The prototype conductors for the Central Solenoid (CS) coils in the CFETR, for example, are designed to produce a peak field of 19.9 T and are expected to be made of high current density Nb 3 Sn strands. Investigations are also ongoing on the application of bismuth strontium calcium copper oxide (BSCCO) and MgB 2 strands for CICCs in fusion reactors. The latter material, MgB 2, could be applied for superconductors subjected to lower magnetic fields, such as Poloidal Field coils, Correction Coils, and Feeders. The performance of all these strands is sensitive to strain, and the mechanical strength of the brittle filaments is relatively weak. This requires a thorough analysis of the cable pattern in terms of the mechanical support of the strands along their length in combination with the minimization of the interstrand coupling currents and strand indentation. As an initial step to finding the most appropriate cable pattern for CICCs, three prototype CICCs made of ITER type Nb 3 Sn strands with significantly different cable twist patterns are tested experimentally for AC coupling loss, interstrand contact resistance, and strand indentation. The three cabling patterns referred to as the Twente, CWS (copper wound superconducting strand), and CFETR-CSMC (CFETR Central Solenoid Model Coil) design. The numerical code JackPot ACDC developed at the University of Twente is used to analyze the interstrand coupling loss and contact resistance. The new ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences) triplet modified CWS design is aimed at reducing strand pinching during cabling, which causes degradation of transport properties during compaction and cyclic loading. The Twente design has the same objective but also aims at reducing the coupling loss while maximizing the mechanical lateral support for the strands by making the twist pitch ratio of the sequential cabling stages close to one. The CFETR-CSMC, taken as a reference for comparison, has cable a pattern mostly similar to the ITER CS cable design.