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.

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.

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 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 detemined 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 sized 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 lage 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 therefor,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.

磁约束聚变堆中的润滑研究

核聚变反应可以提供近乎无限的能源,且是1种安全清洁的方式.磁约束核聚变通过强磁场约束高温高密度氘氚等离子体放电发生聚变反应产生能源,其装备结构复杂且服役环境恶劣.服役于磁约束聚变装置中的活动部件使用的润滑材料不仅面临着摩擦磨损还要承受聚变装置高低温、真空以及辐射环境.本文中系统地总结了磁约束聚变装置发展过程中真空室、超导磁场、离子回旋加热系统和遥操作系统等各个部件面临的润滑挑战以及解决方案,并对未来商用磁约束聚变堆所需润滑材料与技术进行了展望.

基于积分分析方法对聚变堆启动氚量与所需氚增殖比的评估

托卡马克聚变堆的主要发展方式包括混合堆、纯聚变堆。关于托卡马克聚变堆氚自持的研究,国内外主要采用平均滞留时间方法进行研究,并且针对聚变功率较低的混合堆的氚自持研究较少。本工作采用更符合实际的积分分析方法分析了混合堆、纯聚变堆氚自持的启动氚量、氚增殖比(TBR)要求。研究结果表明:启动氚量、备用氚量与聚变功率具有线性关系,所需TBR与聚变功率呈反比例关系;混合堆聚变功率较低,所需TBR较高,工程实现所需TBR挑战较大,需要通过限制长期氚滞留量以降低所需TBR要求;纯聚变堆聚变功率高,所需TBR较低,工程实现所需TBR挑战较小,但备用氚需求达数十千克,应考虑氚系统的冗余设计或提高氚系统的可靠性、可维护性,以降低备用氚的使用规模;运行因子是聚变堆的一个重要设计指标,在此着重分析了运行因子对所需TBR的影响,并重新定义了一个聚变堆氚自持的关系式,以突出运行因子对氚自持的重要影响。

基于cosRMC的聚变堆输运-活化内耦合方法研究

在聚变反应堆运行过程中,中子活化会导致大量放射性核素的产生,因此活化计算在反应堆的屏蔽设计和辐射安全分析中具有重要作用。本文基于蒙特卡罗粒子输运程序cosRMC的内置燃耗求解器Depth,开发了固定源模式下的内耦合输运-活化耦合计算功能。将该程序应用于中国聚变工程试验堆(Chinese Fusion EngineeringTestingReactor,CFETR)的第一壁(First-wall,FW)材料钢和面向等离子体部件(Plasma-facing Component,PFC)材料钨,分别使用连续能量截面和多群截面进行活化计算,并与活化程序ALARA进行对比验证,发现cosRMC的计算结果与ALARA计算的结果符合良好,初步验证了开发的cosRMC程序输运-活化内耦合计算功能的正确性。与传统蒙特卡罗-活化计算程序的外耦合方式相比,内耦合方式蒙特卡罗不需要把中子能谱传给外部活化程序,而是在蒙特卡罗中子输运过程中嵌入计算活化相关的核素单群反应截面,动态更新中子能谱和材料信息,同时可以使用连续能量截面进行反应率计算,得到与实际问题的几何、能谱相关的反应截面,从而精确地考虑了共振区核截面的影响。

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.

聚变堆气态氚排放的辐射影响

聚变堆贮存及释放的气态氚的量远高于现行的裂变堆,氚是聚变堆潜在放射性的重要来源。为未来实现聚变堆的安全性及环境友好性,需要研究聚变堆气态氚排放对环境的影响。选取我国东部沿海典型厂址作为研究对象,使用高斯模型预测气态氚释放后的大气弥散规律以及氚气(HT)的干沉积、土壤氧化以及氚化水(HTO)的再蒸发效应,计算了聚变堆1 g的HT在短期释放情况下对周围环境的公众所造成辐射剂量。计算结果显示:在10 m高度处释放的HT对排放点西部方位500~3 000 m处的成人造成的吸入内照射剂量在0.38~0.10 mSv之间,不同距离HTO的再蒸发效应所造成的剂量都是气态氚剂量的主要来源,沉降至土壤中的HT被氧化成HTO的比例及气象条件是决定气态氚剂量的关键参数。研究表明,聚变堆HT释放所造成公众剂量要高于现行的裂变堆,在后续开展聚变堆的相关研究过程中,需要进一步关注其释放的气态氚对环境的辐射效应。