Fusion Power Generation Research Articles

Loss of power accident analysis for the HCCR blanket

The blanket system is a crucial element in nuclear fusion power generation, tasked with tritium production and the efficient dissipation of the high thermal flux from the tokamak. Korea is in the process of developing the Helium Cooled Ceramic Reflector (HCCR) blanket, which stands as the forefront design among the nation's blanket systems. This paper presents accident analysis results of LOOP (Loss Of Offsite Power) accident for the HCCR blanket. Loss of power accident is one of the postulated events in the nuclear fusion system. This event may be caused by common failure of the external electrical grids or electrical faults in the site. The objective of this safety analysis is to assess the impact of the activation of an additional power supply within 90 seconds after an accident on the system temperature, and to evaluate the heat load that can be removed by the stagnated coolant in the CCWS (Component Cooling Water System) connected to the heat exchanger of the blanket system.

An Improved Voltage Control Method for EAST Fast Control Power Supply

The Experimental Advanced Superconducting Tokamak (EAST) fast control power supply (FCPS) is an important device for controlling the vertical displacement of plasma during the nuclear fusion power generation process, adopting a multiple H-bridge invertor branch parallel operation structure to output total current. At the beginning of each shot of plasma discharge, FCPS works in open-loop voltage control mode (VCM) or closed-loop current control mode (CCM) determined by the plasma control system to output current for exciting the load coil, to achieve plasma vertical displacement control. VCM has the characteristics of fast dynamic response speed but poor consistency of branch current and insufficient branch current control accuracy and stability because of open-loop control. CCM has the characteristics of high branch current control accuracy but poor dynamic response and robustness because of control delay and control parameters determined based on engineering experience. To achieve fast and robust control, an improved voltage control method (IVCM) is proposed by combining the advantages of VCM and CCM. In the beginning of establishing the output current, FCPS operates in VCM, and rapid establishment of the output current is ensured. After the output current rapidly increases to the critical value, closed-loop current control is added to VCM to ensure the accuracy of output current control. In closed-loop current control, linear super-twisting sliding mode control is designed to achieve fast and robust control, ensuring good consistency and fast dynamic response performance of each branch current. Simulations and experiments verify that the designed IVCM has better compatibility characteristics in output current stability, control accuracy, and consistency of each branch current compared to VCM.

Nanotechnology in Nuclear Reactors: Innovations in Fusion and Fission Power Generation

Nanotechnology in Nuclear Reactors: Innovations in Fusion and Fission Power Generation

Mechanism of Fusion Power Generation and Current Status of R&D

Mechanism of Fusion Power Generation and Current Status of R&D

Dawn of clean energy: Enhanced heat transfer, radiative cooling, and firecracker-style controlled nuclear fusion power generation system

Global climate change has become a major environmental threat and development challenge facing humanity. Controllable nuclear fusion is a globally recognized ideal solution for clean energy, but its required high-energy triggering conditions and intense energy release prevent existing technologies from achieving safe, stable, and long-term continuous operation. Here, inspired by the traditional Chinese firecrackers, we propose a pulsed fusion reaction flywheel energy storage multi-reactor relay operation to drive the steam turbine to continuously and stably generate electricity for a long period of time; meanwhile, to install cleaning rotors in the cooling medium pipeline to enhance heat exchange, and to apply radiative cooling technology on the surface of the cooling tower to improve cooling efficiency and to reduce energy consumption, thereby improving system safety and overall energy efficiency. Proposing the combination of original technologies at both the hot end and the cold end of the system, we strive to open up a new way for controllable nuclear fusion power generation.

Tritiated water removal method based on hydrate formation using heavy water as co-precipitant

A separation method of tritium has been developed based on co-precipitation of tritiated water on porous heavy water (D2O) based clathrate hydrate. The tritium was semi-continuously removed by a circulating flow of tritium-containing light water through the D2O hydrate media crystallized in advance. The tritium concentration of 4.77 × 105 Bq/kg was reduced to 1.55 × 103 Bq/kg with HFC-134a hydrate at 0.35 MPa for 1 h, while tritiated water treatment at a concentration of 1010 Bq/kg or greater is eligible with the conventional detritiation methods. Near the D2O hydrate surfaces, solid solutions with higher concentrations of tritiated water were deposited on the hydrate substrate than those in the liquid phase. The semi-serialized operation of hydrate based detritiation is capable of high-rate processing as well as energy and resource saving. This method could be a candidate treatment for huge quantities of water with low concentration tritium discharged from nuclear fusion power generation system.

UNITY: Kyoto Fusioneering’s Unique Integrated Testing Facility for Fusion Power Generation

One of the major research and development challenges on the critical path to achieving a fusion pilot plant is demonstrating the viability of power production from fusion. In August 2022, Kyoto Fusioneering launched the construction of the world’s first integrated testing facility for fusion power generation. The primary testing loop of the facility will initially adopt PbLi, with the prospect of adding an additional molten salt loop, with dimensions of around 5 × 5 × 3 m with a coolant inventory of approximately 100 L. To be completed in 2025, this facility will demonstrate (1) heat extraction from a mock-up blanket, (2) high-temperature heat transfer and exchange, (3) electricity generation from blanket heat, and (4) hydrogen isotope extraction under commercially relevant conditions—to be completed in time for the first-generation demonstration fusion plants. This facility is named UNITY, standing for Unique Integrated Testing facilitY to symbolize the integration of all critical components.

Completion of design and manufacturing of the iter vacuum vessel lower port extension

The ITER is an international project that aims to demonstrate fusion power generation through scientific and engineering research. Most of the major components consisting of the Tokamak machine have been under manufacturing and/or achieved completion of manufacturing with verification of functional tests to be installed as a part of Tokamak. Korean Domestic Agency completed manufacturing of the first ITER VV lower port extension, with the implementation of major technical resolutions, and the rest of units are under fabrication in 2022. There were several technical considerations when conducting the first-of-a-kind design and manufacturing of the ITER VV port component, which will respect the RCC-MR, the essential safety and radioprotection requirements of the French Regulation ESPN. This study explains these technical problems and the solutions established during fabrication.

Analyses and optimization of the CFETR power conversion system with a new supercritical CO2 Brayton cycle

In this paper, the Chinese Fusion Engineering Testing Reactor (CFETR) power conversion system, with a supercritical CO2 (SCO2) Brayton cycle, is designed, analyzed and optimized. Considering the pulse operation of the reactor, a heat storage loop with high temperature molten salt and low temperature concrete is introduced. Based on the parameters of the first cooling loop, the CFETR power conversion loop is designed and studied. A new SCO2 Brayton cycle for the CFETR dual heat sources, blanket and divertor, is developed and optimized using a genetic algorithm. Compared to other simple and recompression cycles, it is shown that the new SCO2 Brayton cycle combines maximum thermal efficiency with simplicity. Exergy analyses are carried out and show that the exergy destruction rates of turbine and heat exchangers between different loops are the largest due to the large turbine power and the large temperature difference. The exergoeconomic analyses show that the fusion reactor accounts for the main cost, which is the key to the economy of fusion power generation. The following sensitivity analyses show that the hot molten salt temperature has a major influence on the system performance. Finally, several multi-criteria optimization algorithms are introduced to simultaneously optimize the three fitness functions, the cycle thermal efficiency, the system exergy efficiency and the total system product unit cost. It is found that the maximum thermal efficiency, the maximum exergy efficiency and the lowest total system product unit cost can be obtained almost simultaneously for the new CFETR power conversion system, and this optimal operation scheme is presented.

The Princeton Field-Reversed Configuration for Compact Nuclear Fusion Power Plants

The Princeton Field-Reversed Configuration (PFRC) nuclear fusion reactor concept is an innovative approach to fusion power generation prioritizing low neutron production and small size. A combination of analytical modeling and numerical simulation shows that the novel heating approach generates an FRC with closed field lines. Simulation data from a single-particle Hamiltonian code predicts ms-scale plasma heating in reactor-scale conditions while PIC codes predict formation of warm FRC plasmas from initial mirror fields. The PFRC-1 and PFRC-2 experiments have heated electrons to energies well in excess of 100 eV and plasma durations to 300 ms, more than 10\(^4\) times longer than the predicted tilt instability growth time. From these data, we have created a development plan and anticipated performance metrics for a fusion reactor based on the PFRC concept. The resulting 1–10 MW PFRC reactors would be suitable for diverse applications, from submarines to urban environments to space propulsion. PFRC is a steady-state, driven magnetic confinement device. Plasma, inside a cylindrical array of coils, is confined and heated by external RF antennae. PFRC would be ultra-low radiation due to both its fuel and small size. The choice of advanced fuels, deuterium and helium-3 (D–3He), may be enabled by the high-\(\beta\) FRC configuration. The small size of the reactor would enable rapid exhaust of the dangerous tritium ash. Low radiation would make the reactor safer to operate and, in combination with simple geometry and small size, dramatically lowers development and maintenance costs. This review paper gives an introduction to the physics of the PFRC and a summary of the PFRC-2 experiment results to date. It then discusses the future program plan and how PFRC reactors would be commercialized.

Towards automated gas leak detection through cluster analysis of mass spectrometer data

In order to generate high-performance plasma for future fusion power generation, it is desirable to keep high quality vacuum during experiment. Mass spectrometer is commonly used to monitor the vacuum quality and to record the amount of atoms and molecules in the vacuum vessel. Leak is the most serious accident to avoid that can nullify an experiment and even harm researchers. Detecting leaks are ever more important since it can be easily overlooked, e.g., when the deterioration in the vacuum degree is modest. This forces the researcher to carefully observe the vacuum and mass spectrometer data. This article presents a way to suggest potential leaks in the vacuum vessel by analyzing mass spectrometer data. This is done by utilizing the Euclidean distance between composition ratios at different times for the clustering using the daily composition ratio. We show that our cluster analysis is an effective way of separating these two cases, which results in a semi-automatic determination of leaks is more efficient than the current norm, which is to check many measures to find a small abnormality in the data manually. We plan further model improvements for long-term evaluation.

Application of Dual Frequency Comb Method as an Approach to Improve the Performance of Multi-Frequency Simultaneous Radiation Doppler Radar for High Temperature Plasma Diagnostics

A new Doppler radar using millimeter-waves in the Ka-band, named the “dual-comb Doppler reflectometer”, has been developed to measure the turbulence intensity and its velocity in high-temperature plasmas. For the realization of a fusion power generation, it is very important to know the spatial structure of turbulence, which is the cause of plasma confinement degradation. As a non-invasive and high spatial resolution measurement method for this purpose, we apply a multi-frequency Doppler radar especially with simultaneous multi-point measurement using a frequency comb. The newly developed method of synchronizing two frequency combs allows a lower intermediate frequency (IF) than the previously developed frequency comb radar, lowering the bandwidth of the data acquisition system and enabling low-cost, long-duration plasma measurements. In the current dual-comb radar system, IF bandwidth is less than 0.5 GHz; it used to be 8 GHz for the entire Ka-band probing. We applied this system to the high-temperature plasma experimental device, the Large Helical Device (LHD), and, to demonstrate this system, verified that it shows time variation similar to that of the existing Doppler radar measurements.

Deconvolving the roles of E × B shear and pedestal structure in the energy confinement quality of super H-mode experiments

Analysis of ‘super H-mode’ experiments on DIII-D has put forward that high plasma toroidal rotation, not high pedestal, plays the essential role in achieving energy confinement quality H 98y2 ≫ 1 (Ding et al 2020 Nucl. Fusion 60 034001). Recently, super H-mode experiments with variable input torque have confirmed that high rotation shear discharges have very high levels of H 98y2 (>1.5), independent of the pedestal height, and that high pedestal discharges with low rotation shear have levels of H 98y2 only slightly above 1 (⩽1.2). Although some increase in stored energy with higher pedestal occurs, the energy confinement quality mainly depends on the toroidal rotation shear, which varies according to different levels of injected neutral beam torque per particle. Quasi-linear gyrofluid modeling achieves a good match of the experiment when including the E × B shear; without including plasma rotation, the modeling predicts a confinement quality consistent with the empirical observation of H 98y2 ∼ 1.2 at low rotation. Nonlinear gyrokinetic transport modeling shows that the effect of E × B turbulence stabilization is far larger than other mechanisms, such as the so-called hot-ion stabilization (T i/T e) effect. Consistent with these experimental and modeling results are previous simulations of the ITER baseline scenario using a super H-mode pedestal solution (Solomon et al 2016 Phys. Plasmas 23 056105), which showed the potential to exceed the Q = 10 target if the pedestal density could be increased above the Greenwald limit. A close look at these simulations reveals that the predicted energy confinement quality is below 1 even at the highest pedestal pressure. The improvement in Q at higher pedestal density is due to the improved fusion power generation at the higher core density associated with higher pedestal density, not to an improved energy confinement quality.

Core size effects of laser fusion subcritical research reactor for fusion engineering research

A multi-purpose high repetition laser facility, the so-called Japan establishment for power-laser community harvest (J-EPoCH) is proposed as a next generation laser facility. J-EPoCH will operate at the maximum rate of 100 Hz. The omnidirectional 12 laser beams with 8 kJ would yield ∼1013 neutrons with a large high aspect ratio target. As one of the applications of J-EPoCH, a laser fusion subcritical research reactor has been conceptually designed based on existing technologies. Moreover, a variety of fusion engineering studies: energy conversion, tritium breeding, neutron irradiation effects, etc, can be conducted. The feasibility of the subcritical research reactor is considered in terms of neutron-thermal (n-t) conversion and tritium breeding. Lead–lithium alloy (Li17Pb83) and boron carbide (B4C) have the potential to be studied for preliminary fusion power generation. The subcritical reactor will generate 21.4 W and 20.0 W of the thermal fusion power with the Li17Pb83 and the B4C layers of the thickness of 80 cm, respectively at 1 Hz operation. The Li17Pb83 layer of a 5 mm thickness will achieve the temperature rise of 0.203 mK per shot. The thermal fusion energy is detectable with conventional measurement techniques. The core with the Li17Pb83 layer thickness of 100 cm will yield more than one tritium from a deuterium–tritium fusion neutron. However, laser windows reduce the efficiency of n-t conversion and tritium yield.

RAMI analysis for the ITER In-Vessel Coils System

The ITER project is the largest international collaboration in the scientific field ever set up. It forms the heart of a unique collaborative agreement to build the first experimental nuclear fusion device designed to prove the scientific and technological feasibility of sustained fusion power generation. Among hundreds of systems, the In-Vessel Coils System plays a crucial role in ITER: ensuring and maintaining stable plasma operations. ITER has an ambitious inherent availability target for plasma production. To be able to achieve this objective, the potential technical risk that may affect the machine operation, was assessed by means of RAMI (Reliability, Availability, Maintainability and Inspectability) approach. The RAMI analysis for the IVC system was performed on the system design, following the Staged Approach Configuration adopted by ITER for Construction and Operation phases. A functional breakdown was prepared, resulting in the system being divided into 3 main functions described using the IDEFØ method. A qualitative analysis was performed applying the Failure Modes, Effects and Critical Analysis technique in order to identify the potential effects due to different anomalies of the IVC components. Criticality chart highlights the distribution of the Failure Modes across three risk zones with regard to their probability of occurrence and the impact on the availability. In relation to the operation of the IVC System, the analysis identified 32 risks, none of whom were classified as major. To estimate the reliability and availability of IVC system and to assess the compliance to the requirements, a quantitative analysis was performed using the Reliability Block Diagrams Analysis. The inherent availability of the IVC system during DT phase was calculated over the 20 years of ITER lifetime. The availability of the ITER IVC system results to be compliant with the inherent availability requirement for DT phase.

Power exhaust by core radiation at COMPASS tokamak

Substantial power dissipation in the edge plasma is required for the safe operation of ITER and next-step fusion reactors, otherwise unmitigated heat fluxes at the divertor plasma-facing components (PFCs) would easily exceed their material limits. Traditionally, such heat flux mitigation is linked to the regime of detachment, which is characterised by a significant pressure gradient between upstream and downstream scrape-off layer (SOL). However, the physics phenomena responsible for power dissipation and pressure loss are distinctly different, especially when the power dissipation is achieved by impurity seeding. In principle, it is possible to achieve substantial mitigation of the heat fluxes while maintaining conservation of the pressure along the open field lines in the SOL. This regime can be accessed by injection of medium- or high-Z impurities, which mostly radiate inside the last closed flux surface. The critical question related to such an approach is the effect on confinement and perspective fusion power generation in future thermonuclear reactors. In this work, we report on experiments at COMPASS tokamak, where neon and argon impurities were injected in ohmic or NBI-heated low confinement plasmas. With appropriate seeding waveform, stable scenarios were achieved, avoiding the radiative collapse of plasmas. Significant reduction of heat fluxes at the outer target was observed, with heat flux pattern similar to the one previously achieved by nitrogen seeding. The reduction of downstream pressure was, however, accompanied by an equal reduction of upstream pressure, indicating that the power dissipation occurred inside the separatrix. Indeed, the impurity cooling is causing a significant drop of edge temperature; however, the effect in the plasma centre is much less pronounced.

Conceptual Design of Isolated Power for Quench Detection System With Highly Insulating Stability Under Super High Field Magnets Using Wireless Power Transfer Technology

The super high field magnets using superconducting wires have been promisingly applied for thermonuclear fusion power generation such as Korea Superconducting Tokamak Advanced Research (KSTAR) and International Thermonuclear experimental reactor (ITER) systems. The quench detection system (QDS) is essential for high reliability in the super high field magnet facilities. Generally, the quench voltage detection system, which uses resistive voltages, has been adopted. Such an ITER magnet, the maximum operation current and voltage of superconducting keeps over 60kA and 50 kV DC. Generally, in the case of medium power, the power transformer has been adopted for QDS as an isolated power due to stable insulating and durable characteristics. However, unfortunately, the magnetic saturation of electric steel for power transformer remained fragile as well as the volume is increased under super high field magnet. From this reason, authors suggest the electron quench detection device (EQDD) module. The main functions of EQDD are to convert high voltage signals of magnet into voltage-drop, and then send into quench monitoring system through optic cable. The EQDD module needs to isolate power system since EQDD module should be operated under high voltage insulation. From this reason, authors apply the wireless power system for EQDD module as a reasonable option instead of power transformer since the wireless power system can transfer power through any non-metallic media between resonance antenna (Tx) and receiver (Rx) coils. In this study, authors describe the possibility of conceptual design for quench detection system with high insulation under super high field magnet using wireless power transfer technology. As the EQDD including isolated wireless power should be electrically shielded, the shielding effects for Tx coil is investigated under different intervals of shielding barrier. Additionally, to expand the delivery distance between Tx and Rx coils, the operation and thermal characteristics for transfer efficiency are investigated under 370 kHz and 750 kHz RF generators, respectively.

Potential contribution of fusion power generation to low-carbon development under the Paris Agreement and associated uncertainties

Fusion energy has abundant resources, does not exhibit nuclear runaway, and it has zero-carbon emissions. Long-term strategies for fusion energy development will become critically important to promote future DEMOnstration Fusion Reactor (DEMO) projects by another largescale investment and gain social acceptance. This paper assessed the potential contribution of fusion power generation to low-carbon development, which is prescribed in the Paris Agreement, under the combination of different uncertainties of future socioeconomic development, probability of the 2 °C target, and development of commercial fusion power plants. Global negative CO2 emission in 2100 by drastic decarbonization of energy systems was required to achieve the 2 °C target, and fusion power plants were expected to be installed in the latter half of the 21st century mainly in countries with limited potentials of zero-emission energy sources, such as Japan, Korea, and Turkey, for cost-efficient climate change mitigation. If inexpensive power plants could be developed by enhanced R&D and advanced design in DEMO projects or if the establishment of fission plants in the future is low, fusion power generation will also be deployed in the EU28, India, and China. Further cost reduction by innovative design and alternative concepts are also essential to diffuse fusion plants in zero-emission resource-rich countries.

The Plasma Compression Fusion Device—Enabling Nuclear Fusion Ignition

The plasma compression fusion device (PCFD) generates the energy gain by plasma compression-induced nuclear fusion. This concept has the capability of maximizing the product of plasma pressure and energy confinement time to maximize the energy gain, and thus give rise to fusion ignition conditions. The preferred embodiment of this original concept uses a hollow cross-duct configuration of circular cross section in which the concentrated magnetic energy flux from two pairs of opposing curved-headed counter-spinning conical structures (possibly made from an alloy of tungsten with high capacitance) whose outer surfaces are electrically charged compresses a gaseous mixture of fusion fuel into a plasma, heated to extreme temperatures and pressures. The generated high-intensity electromagnetic (EM) radiation heats the plasma and the produced magnetic fields confine it in between the counter-spinning conical structures, named the dynamic fusors (four of them-smoothly curved apex sections opposing each other in pairs). The dynamic fusors can be assemblies of electrified grids and toroidal magnetic coils, arranged within a conical structure whose outer surface is electrically charged. The cross-duct inner surface surrounding the plasma core region is also electrically charged and vibrated in an accelerated mode to minimize the flux of plasma particles (including neutrals) from impacting the PCFD surfaces and initiating a plasma quench. The fusion fuel (preferably deuterium gas) is introduced into the plasma core through the counterspinning conical structures, namely, injected through orifices in the dynamic fusor heads. There is envisioned another even more compact version of this concept, which uses accelerated vibration in a linear-duct configuration (using two counterspinning dynamic fusors only) and would best be suited for fusion power generation on aircraft, or main battle tanks. The concept uses controlled motion of electrically charged matter through accelerated vibration and/or accelerated spin subjected to smooth, yet rapid acceleration transients, to generate extremely high-energy/high-intensity EM radiation (fields of high-energy photons) which not only confines the plasma but also greatly compresses itso as to produce a high power density plasma burn, leading to ignition. The PCFD concept can produce power in the gigawatt to terawatt range (and higher) with input power in the kilowatt to megawatt range and can possibly lead to ignition (selfsustained) plasma burn. Several important practical engineering and operational issues with operating a device such as the PCFD are discussed.

Design and Development of Mirnov Coil Sensor for Eddy Currents Experiment on Toroidal Vessel

Tokamak is a magnetic confinement device that confines hot plasma in the shape of torus during the process of thermonuclear fusion power generation. In tokamak, eddy currents are produced due to change in plasma positions during plasma instabilities that induce electromagnetic forces on interaction with the induced currents. Mirnov coils are widely used in tokamaks to study plasma positions during plasma instabilities. Principle objective of this paper is the design and development of Mirnov coil sensor to find eddy currents on a toroidal vessel. This paper presents an elaborative and practical construction technique of a Mirnov coil. The calibration method of a Mirnov coil is also discussed. Mirnov coils as an eddy current diagnostics are tested and experiments to measure magnetic fields due to induced current on torroidal vessel are performed using the coils.