An innovative approach to effective breeding blanket design for future fusion reactors

An effective breeding blanket is critical to support tritium self-sufficiency for future fusion reactors.The difficulty is to achieve tritium breeding ratio(TBR)target of 1.05 or more.This paper presents a new design approach to the blanket design process.It indicates that fusion blanket design is affected by universal functions based on iterations.Three aspects are worth more attention from fusion engineers in the future.The first factor is that the iterations on the material fractions affect not only structure scheme but also TBR variation.The second factor is the cooling condition affecting final TBR due to the change of the structure material proportion.The third factor is temperature field related to the tritium release.In particular,it is suggested that the statistical calculation of effective TBR must be under reasonable control of the blanket temperature field.This approach is novel for blanket engineering in development of a fusion reactor.

Proposal of a Deuterium-Deuterium Fusion Reactor Intended for a Large Power Plant

This article looks for the necessary conditions to use Deuterium-Deuterium (D-D) fusion for a large power plant. At the moment, for nearly all the projects (JET, ITER…) only the Deuterium-Tritium (D-T) fuel is considered for a power plant. However, as shown in this article, even if a D-D reactor would be necessarily much bigger than a D-T reactor due to the much weaker fusion reactivity of the D-D fusion compared to the D-T fusion, a D-D reactor size would remain under an acceptable size. Indeed, a D-D power plant would be necessarily large and powerful, i.e. the net electric power would be equal to a minimum of 1.2 GWe and preferably above 10 GWe. A D-D reactor would be less complex than a D-T reactor as it is not necessary to obtain Tritium from the reactor itself. It is proposed the same type of reactor yet proposed by the author in a previous article, i.e. a Stellarator “racetrack” magnetic loop. The working of this reactor is continuous. It is reminded that the Deuterium is relatively abundant on the sea water, and so it constitutes an almost inexhaustible source of energy. Thanks to secondary fusions (D-T and D-He3) which both occur at an appreciable level above 100 keV, plasma can stabilize around such high equilibrium energy (i.e. between 100 and 150 keV). The mechanical gain (Q) of such reactor increases with the internal pipe radius, up to 4.5 m. A radius of 4.5 m permits a mechanical gain (Q) of about 17 which thanks to a modern thermo-dynamical conversion would lead to convert about 21% of the thermal power issued from the D-D reactor in a net electric power of 20 GWe. The goal of the article is to create a physical model of the D-D reactor so as to estimate this one without the need of a simulator and finally to estimate the dimensions, power and yield of such D-D reactor for different net electrical powers. The difficulties of the modeling of such reactor are listed in this article and would certainly be applicable to a future D-He3 reactor, if any.

Induction System for a Fusion Reactor: Quantum Mechanics Chained up

In the quest for a sustainable and abundant energy source, nuclear fusion technology stands as a beacon of hope. This study introduces a groundbreaking quantum mechanically effective induction system designed for magnetic plasma confinement within fusion reactors. The pursuit of clean energy, essential to combat climate change, hinges on the ability to harness nuclear fusion efficiently. Traditional approaches have faced challenges in plasma stability and energy efficiency. The novel induction system presented here not only addresses these issues but also transforms fusion reactors into integrated construction systems. This innovation promises compact fusion reactors, marking a significant step toward a clean and limitless energy future, free from the constraints of traditional power sources. This revolutionary quantum induction system redefines plasma confinement in fusion reactors, unlocking clean, compact, and efficient energy production.

MHD Stability Analysis and Flow Controls of Liquid Metal Free Surface Film Flows as Fusion Reactor PFCs

Numerical and experimental investigation results on the magnetohydrodynamics（MHD） film flows along flat and curved bottom surfaces are summarized in this study. A simplified modeling has been developed to study the liquid metal MHD film state, which has been validated by the existing experimental results. Numerical results on how the inlet velocity（V）, the chute width（W） and the inlet film thickness（d0） affect the MHD film flow state are obtained. MHD stability analysis results are also provided in this study. The results show that strong magnetic fields make the stable V decrease several times compared to the case with no magnetic field,especially small radial magnetic fields（Bn） will have a significant impact on the MHD film flow state. Based on the above numerical and MHD stability analysis results flow control methods are proposed for flat and curved MHD film flows. For curved film flow we firstly proposed a new multi-layers MHD film flow system with a solid metal mesh to get the stable MHD film flows along the curved bottom surface. Experiments on flat and curved MHD film flows are also carried out and some firstly observed results are achieved.

Relationship between Low Temperature Toughness and Microstructure in Low Activation Fe-Cr-Mn(W,V)Steel for Fusion Reactors

Fe-Cr-Mn (W, V) austenite steel was researched in order to supply a theory base for the first wall materials of fusion reactors.Experiments included vacuum melting, forging, annealing, solution treatment, Charpy impact tests and microstructure observation. Theresults show that the imped value decreases with the test temperature decreasing. In this system, there is ductile/brittle transition. Themechanism of this decrease of the impact value is considered to be due to γ - ε transformation in sub-stable austenite steel and stoppingoverlapping sacking fault by grain boundaries in stable austenite steel.

Application of Kelvin Probe to Studies of Fusion Reactor Materials under Irradiation

Recently, the work function （WF） changes in metallic and ceramic materials to be potentially used in future fusion reactors have been examined by means of Kelvin probe （KP）, under He ion irradiation in high energy （MeV） and / or low energy （500 eV） ranges. The results of polycrystalline Ni samples indicate that the 1 MeV beam only induces decrease in the WF within the experimental fluence range; whereas the irradiation of 500 eV beam results in decrease in the WF firstly, then increase till saturation. A dual layer surface model is employed to explain the observed phenomena, together with computer simulation results by SRIM code. Charges buildup on the surface of lithium ceramics has been found to greatly influence the probe output, which can be explained qualitatively using a model concerning an induction electric field due to external field and free charges on the ceramic surface.

Core Plasma Characteristics of a Spherical Tokamak D-^3He Fusion Reactor

The magnetic fusion reactor using the advanced D-3He fuels has the advantage of much less-neutron productions so that the consequent damages to the first wall are less serious. If the establishment of this kind of reactor becomes realistic, the exploration of 3He on the moon will be largely motivated. Based on recent progresses in the spherical torus (ST) research, we have physically designed a D-3He fusion reactor using the extrapolated results from the ST experiments and also the present-day tokamak scaling. It is found that the reactor size significantly depends on the wall reflection coefficient of the synchrotron radiation and of the impurity contaminations. The secondary reaction between D-D that promptly leads to the D-T reaction producing 14 MeV neutrons is also estimated. Comparison of this D-3He ST reactor with the D-T reactor is made.

Progressive Thermalization Fusion Reactor Able to Produce Nuclear Fusions at Higher Mechanical Gain

In the standard fusion reactors, mainly tokamaks, the mechanical gain obtained is below 1. On the other hand, there are colliding beam fusion reactors, for which, the not neutral plasma and the space charge limit the number of fusions to a very small number. Consequently, the mechanical gain is extremely low. The proposed reactor is also a colliding beam fusion reactor, configured in Stellarator, using directed beams. D+/T+ ions are injected in opposition, with electrons, at high speeds, so as to form a neutral beam. All these particles turn in a magnetic loop in form of figure of “0” (“racetrack”). The plasma is initially non-thermal but, as expected, rapidly becomes thermal, so all states between non-thermal and thermal exist in this reactor. The main advantage of this reactor is that this plasma after having been brought up near to the optimum conditions for fusion (around 68 keV), is then maintained in this state, thanks to low energy non-thermal ions (≤15 keV). So the energetic cost is low and the mechanical gain (Q) is high (>>1). The goal of this article is to study a different type of fusion reactor, its advantages (no net plasma current inside this reactor, so no disruptive instabilities and consequently a continuous working, a relatively simple way to control the reactor thanks to the particles injectors), and its drawbacks, using a simulator tool. The finding results are valuable for possible future fusion reactors able to generate massive energy in a cleaner and safer way than fission reactors.

Magnetohydrodynamic Calculations of Toroidal Fusion Reactor to Ensure Stable Control

The development of magnetic configurations to confine the stability fluid plasmas for fusion energy is a challenge that is a mixture of basic fusion engineering and invention. In order to keep the fusion reactions in the plasma to be continuing in the fusion reactors, the speed of tritium breeding (TBR) should be kept above a certain value. At the Apex fusion reactor, a fast flowing thin liquid wall has replaced the solid first wall concept of the traditional reactors. Behind the fast flowing thin liquid wall, a slower and thicker second liquid wall (coat) is present. Monte Carlo Random method (MCRS) is the general name for the solution of experimental and statistical problems with a random approach. This method is dependent upon the theory of probability. In the present work, Mhd impacts are investigated quite unimportant for Flibe salt solutions. In this study, the fissile fuel production calculations are done for a neutron wall load of 10 MW/m2 fissile fuel production rates of 238U(n, γ)239Pu and 232Th(n,γ)233U increases almost linearly with increased heavy metal content.

Overview of Fusion Reactor Design

In this paper, the progress of the research activities of fusion reactor design at SWIP in 2006 is introduced.

Plasma Theory——Possibility of Establishing D-^3He Fusion Reactor Using Spherical Tokamaks

In recent years, a new approach named the spherical tokamak or spherical torus （ST） in the magnetic fusion research has made remarkable progress, parallel to the tokamak development including the international thermonuclear experimental reactor （ITER） projectTM. In ST experiments, magnetohydrodynamics stable high beta value （the ratio of the plasma pressure to the toroidal magnetic pressure） up to 50% has been routinely obtained. The confinement scaling for ST, though being less-confident compared to the database of tokamaks, seems at least to be as good as the tokamak.

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

Activation analyses play a vital role in nuclear reactor design.Activation analyses,along with nuclear analyses,provide important information for nuclear safety and maintenance strategies.Activation analyses also help in the selection of materials for a nuclear reactor,by providing the radioactivity and dose rate levels after irradiation.This information is important to help define maintenance activity for different parts of the reactor,and to plan decommissioning and radioactive waste disposal strategies.The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential,due to the presence of a highenergy neutron environment which makes decisive demands on material selection.This study comprises two parts; in the first part the activation characteristics,in a fusion radiation environment,of several elements which are widely present in structural materials,are studied.It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment.The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions.The structural materials selected for this study,i.e.India-specific Reduced Activation Ferritic-Martensitic steel（IN-RAFMS）,P91-grade steel,stainless steel 316 LN ITER-grade（SS-316 LN-IG）,stainless steel 316 L and stainless steel 304,are candidates for use in ITER either in vessel components or test blanket systems.Tungsten is also included in this study because of its use for ITER plasma-facing components.The study is carried out using the reference parameters of the ITER fusion reactor.The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port.The presence of elements like Nb,Mo,Co and Ta in a structural material enhance the activity level as well as the dose level,which has an impact on design considerations.IN-RAFMS was shown to be a more effective low-activation material than SS-316 LN-IG.

Estimates of Tritium Produced Ratio in the Blanket of Fusion Reactors

For the preparation of tritium fuel as the main and rare fuel of reactors in the fusion reactors, the reactor blanket must be designed so that it provides enough tritium breeding ratio. The tritium breeding ratio, TBR, in the blanket of reactors should be greater than one, (TBR > 1), by applying lithium blanket. The calculations for proposed parameters (td , fb , η and tp), indicate that the estimated tritium breeding ratio is greater than one. The calculated TBR = 1.04 satisfies the tritium provision condition.

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

This article explores the transformative potential of nanotechnology and MMs(memory metals)in enhancing the design and operation of nuclear reactors,encompassing both fission and fusion technologies.Nanotechnology,with its ability to engineer materials at the atomic scale,offers significant improvements in reactor safety,efficiency,and longevity.In fission reactors,nanomaterials enhance fuel rod integrity,optimize thermal management,and improve in-core instrumentation.Fusion reactors benefit from nanostructured materials that bolster containment and heat dissipation,addressing critical challenges in sustaining fusion reactions.The integration of SMAs(shape memory alloys),or MMs,further amplifies these advancements.These materials,characterized by their ability to revert to a pre-defined shape under thermal conditions,provide self-healing capabilities,adaptive structural components,and enhanced magnetic confinement.The synergy between nanotechnology and MMs represents a paradigm shift in nuclear reactor technology,promising a future of cleaner,more efficient,and safer nuclear energy production.This innovative approach positions the nuclear industry to meet the growing global energy demand while addressing environmental and safety concerns.

Proposal of a Deuterium-Deuterium Fusion/PWR Fission Hybrid Reactor

This article proposes to associate a Deuterium-Deuterium (D-D) fusion reactor with a PWR (fission Pressurized Water Reactor) in a hybrid reactor. Even if the mechanical gain (Q factor) of the D-D fusion reactor is below the unity and consequently consumes more energy than it supplies, due to the high energy amplification factor of the PWR fission reactor, the global yield is widely superior to 1. As the energy supplied by the fusion reactor is relatively low and as the neutrons supplied are mainly issued from D-D fusions (at 2.45 MeV), the problems of heat flux and neutrons damage connected with materials, as with D-T fusion reactors are reduced. Of course, there is no need to produce Tritium with this D-D fusion reactor. This type of reactor is able to incinerate any mixture of natural Uranium, natural Thorium and depleted Uranium (waste issued from enrichment plants), with natural Thorium being the best choice. No enriched fuel is needed. So, this type of reactor could constitute a source of energy for several thousands of years because it is about 90 more efficient than a standard fission reactor, such as a PWR or a Candu one, by extracting almost completely the energy from the fertile materials U238 and Th232. For the fission part, PWR technology is mature. For the fusion part, it is based on a reasonable hypothesis done on present Stellarators projects. The working of this reactor is continuous, 24 hours a day. In this paper, it will be targeted a reactor able to provide net electric power of about 1400 MWe, as a big fission power plant.

Review on development of reduced activated ferritic/martensitic steel for fusion reactor

Materials are stil one of the main technical bottlenecks restricting the development of fusion reactors.Reduced activated ferritic/martensitic steel(RAFM)is considered one of the main candidate structural materials for fusion reactor cladding due to its good radiation resistance and mechanical properties.In the past 40 years,RAFM steel has made considerable progress,but numerous problems remain to be solved.The improvements in RAFM steel in recent years,such as chemical composition optimization,clean preparation technology,radiation performance,and applicable welding technology,were systematically summarized.A systematic review of new RAFM steels was conducted,the development direction of the traditional smelting process was analyzed,and the application of laser additive manufacturing technology to RAFM steel was introduced.The effect of rradiation on the microstructure and mechanical properties of RAFM steel was described,and welding methods of RAFM steel and their research progress were reviewed.Finally,the potential applications of Si,Ti,and Zr in improving the performance of RAFM steel,electroslag remelting technology in clean smelting,heat treatment process in optimizing radiation performance,and laser-beam welding in RAFM welding were prospected and summarized.

Fundamental Analysis of Helium-Gas Coolant Leakage Rate Through First-Wall Cracks in Tokamak Fusion Reactors

A fundamental analysis of helium-gas coolant leakage rate through first-wall cracks in Tokamak fusion reactors was made. Criteria for ascertaining the correct flow models were thoroughly investigated. After testing the criteria, it was determined that the correct model is the compressible choked flow for the helium-gas coolant under the normal operating conditions in the Tokamak fusion reactors. The upper bound leakage rates through metallic wall for two crack sizes were calculated. The calculated maximum numbers of allowable cracks through metallic and silicon-carbon composite wall were also reported. The experimental data of specimen S-23 (the small crack size), checked with the predicted or calculated leakage rate. But the experimental data of specimen S-4 (the large crack size, which is only 4.4 times larger than the crack size of specimen S-23) were two orders of magnitude higher than the calculated value. This is probably due to the many through-cracks undetected and therefore, not reported in the experiment, and not due to the difference in crack sizes. It should be noted that since there are only two test data points, it is recommended that more testing or experimental data will be needed. The results of two previous investigations about the calculated leakage values, their equations used, and their flow models employed were also reviewed. It is concluded that the correct model for the analysis is the compressible choked flow, and that helium can be as an effective coolant for fusion power reactors. Several recommendations are also made. Specifically, more experiments for helium, and similar analysis and experiments for lithium and water coolant are needed; and should be encouraged.

Spark plasma sintering of tungsten-based WTaVCr refractory high entropy alloys for nuclear fusion applications

W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a powder metallurgy process combining mechanical alloying and spark plasma sintering (SPS). The SPSed samples contained two phases, in which the matrix is RHEA with a body-centered cubic structure, while the oxide phase was most likely Ta2VO6through a combined analysis of X-ray diffraction (XRD),energy-dispersive spectroscopy (EDS), and selected area electron diffraction (SAED). The higher oxygen affinity of Ta and V may explain the preferential formation of their oxide phases based on thermodynamic calculations. Electron backscatter diffraction (EBSD) revealed an average grain size of 6.2μm. WTaVCr RHEA showed a peak compressive strength of 2997 MPa at room temperature and much higher micro-and nano-hardness than W and other W-based RHEAs in the literature. Their high Rockwell hardness can be retained to at least 1000°C.

Review of heavy-ion inertial fusion physics

In this review paper on heavy ion inertial fusion(HIF),the state-of-the-art scientific results are presented and discussed on the HIF physics,including physics of the heavy ion beam(HIB)transport in a fusion reactor,the HIBs-ion illumination on a direct-drive fuel target,the fuel target physics,the uniformity of the HIF target implosion,the smoothing mechanisms of the target implosion non-uniformity and the robust target implosion.The HIB has remarkable preferable features to release the fusion energy in inertial fusion:in particle accelerators HIBs are generated with a high driver efficiency of~30%-40%,and the HIB ions deposit their energy inside of materials.Therefore,a requirement for the fusion target energy gain is relatively low,that would be~50-70 to operate a HIF fusion reactor with the standard energy output of 1 GWof electricity.The HIF reactor operation frequency would be~10-15 Hz or so.Several-MJ HIBs illuminate a fusion fuel target,and the fuel target is imploded to about a thousand times of the solid density.Then the DT fuel is ignited and burned.The HIB ion deposition range is defined by the HIB ions stopping length,which would be~1 mm or so depending on the material.Therefore,a relatively large density-scale length appears in the fuel target material.One of the critical issues in inertial fusion would be a spherically uniform target compression,which would be degraded by a non-uniform implosion.The implosion non-uniformity would be introduced by the Rayleigh-Taylor(R-T)instability,and the large densitygradient-scale length helps to reduce the R-T growth rate.On the other hand,the large scale length of the HIB ions stopping range suggests that the temperature at the energy deposition layer in a HIF target does not reach a very-high temperature:normally about 300 eV or so is realized in the energy absorption region,and that a direct-drive target would be appropriate in HIF.In addition,the HIB accelerators are operated repetitively and stably.The precise control of the HIB axis manipulation is also realized in the HIF accelerator,and the HIB wobbling motion may give another tool to smooth the HIB illumination non-uniformity.The key issues in HIF physics are also discussed and presented in the paper.

Fusion Material Studies Relating to Safety in Russia in 2002

The paper is a summary of Russian material studies performed in frames of activi-ties aiming at substantiation of safety of the International Thermonuclear Experimental Reactor(ITER)after 2001.Subthreshold sputtering of tungsten by 5 eV deuterons was revealed at temper-atures above 1150℃.Mechanism of globular films formation was further studied.Computations of tritium permeation into vacuum vessel coolant confirmed the acceptability of vacuum vessel cooling system for removal of the decay heat.The most dangerous accident with high-current are in toroidal superconducting magnets able to burn out a bore up to 0.6 m in diameter in the cryostat vessel was determined.Radiochemical reprocessing of V-Cr-Ti alloy and its purification from activation products down to a contact dose rate of~10μSv/h was developed.