冯开明先进燃料靶惯性约束聚变的燃耗和增益

计算了球形均匀冯开明先进燃料靶惯性约束聚变(ICF)的燃耗和增益。讨论了这种堆系统的能量平衡。设计了一种新型的由毛细管阵列组成具有抗辐射损伤、可自动更新的液悉金属锂自由表面多孔漫璧。用它取出聚变能。同时与D-T热核燃料靶系统的燃耗和增益及它们不同的堆工程特性作了比较。     

聚变堆与聚变堆材料——D-^3He先进燃料靶惯性约束聚变的燃耗和增益

本文研究采用大功率激光束压缩球形均匀D-^3He先进燃料靶聚变，推导和计算了这种惯性约束聚变（ICF）的靶丸燃耗和增益，数值计算表明，只有在小火花内能和较高惯性约束参数HF近似下，对典型的堆系统效率参数，才有可能使D-^3He靶ICF系统有大的净能量输出。相反，在大火花内能和低HF近似下，D-^3He靶ICF系统几乎不可能得到净的能量输出，从而证明了快点火思维方法的科学依据。     

次临界能源堆物理设计进展

聚变-裂变混合能源堆包括聚变中子源和次临界能源堆,主要目标是生产电能。回顾了国内外混合堆的发展历史,给出混合能源堆设计的边界条件和约束条件,说明次临界能源堆以铀锆合金为燃料、水为冷却剂的设计思想。利用输运燃耗耦合程序MCORGS计算了混合能源的燃耗,给出了中子有效增殖因数、能量放大倍数和氚增殖比等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。论述了混合堆的热工设计并进行了安全分析。对于燃耗数值模拟程序,通过多家对算,保证其计算结果的可信性。针对次临界能源堆的特点,利用贫铀球壳建立了贫铀聚乙烯装置和贫铀LiH装置,并且专门设计加工了天然铀装置,开展铀裂变率、造钚率、产氚率等中子学积分实验,验证了数值模拟的可靠性。     

球形托卡马克堆嬗变中子学计算的比较研究

基于对球形托卡马克（ST）聚变堆的研究，提出了ST聚变－嬗变堆的设计概念。运用一维输运燃耗计算程序BISON3.0进行了优化设计，确定了适合于嬗变少额锕系MA核素的堆芯等离子体参数、包层结构及合适的换料周期。在一维计算的基础上，运用二维中子学程序TWODANT进行了二维中子输运计算；结合TWODANT给出的中子通量，运用一维放射性计算程序FDKR进行了燃耗计算，并给出了有关的计算结果。     

聚变裂变混合堆处理高放超铀废物的研究

提出一个燃烧高放超铀废物的思路,即在外部聚变中子源驱动下,把燃烧超铀锕系元素和钍铀燃料循环相结合.并且设计相应的一维模型,使用开发的燃耗计算程序ONESN_BURN和新制作的数据库对模型进行计算和分析.通过计算,得到锕系元素的放射性,生物潜在危害因子,高放超铀锕系废物的密度和非常深的燃耗深度等.比较聚变裂变混合堆与传统的热堆,发现中子能谱越硬,对燃烧超铀锕系元素越有效.     

Z箍缩驱动聚变-裂变混合能源堆总体概念研究

中国工程物理研究院提出的Z箍缩驱动聚变-裂变混合能源堆(Z-FFR)概念,采用Z箍缩热核聚变产生的大量中子驱动次临界裂变堆而释放能量,集成了"局部整体点火"聚变靶、"先进次临界能源堆"等创新概念,在安全、经济、持久和环境友好等方面具有优良的品质,有望成为有效应对未来能源危机和环境气候问题的千年能源。简要回顾了国内外Z箍缩聚变能源(Z-IFE)的相关研究进展,介绍了中国工程物理研究院在Z-FFR方向的总体概念研究情况,从驱动器、聚变靶设计和次临界裂变堆三方面阐述了此能源系统的原理结构和运行特点,对其经济性进行了评估,同时提出了未来Z-FFR的发展路线图设想。     

氘氚聚变中子发生器旋转氚靶传热特性研究

强流氘氚中子发生器可用于模拟聚变堆中子环境, 对于开展聚变堆包层材料相关实验研究具有重要意义. 本文提出了一种用于10¹² n·^-1量级氘氚中子发生器HINEG (high intensity neutron generator)的旋转氚靶系统设计方案, 并对其技术难点和强化传热方法进行了介绍. 为考查该氚靶系统的传热特性, 利用Computational Fluid Dynamics方法对冷却水层厚度、冷却水流速和氚靶系统旋转速度对靶面冷却的影响进行了分析, 并对不同热功率密度下靶面的传热过程进行了研究. 结果显示, 大的水层厚度、大的冷却水流速和高的靶系统旋转速度有利于靶面的冷却, 但水层厚度和水流速的变化对靶面传热影响较小. 一定条件下靶面所承受的热功率密度不能超过某个限值.     

In-Zinerater液态包层输运燃耗数值模拟

Z-Pinch惯性约束聚变是未来一种有竞争力的能源候选方案。Z-Pinch驱动的聚变裂变混合堆可高效地嬗变反应堆乏燃料中分离出的超铀元素。对美国Sandia国家实验室提出的In-Zinerater混合堆概念进行了中子学分析和数值模拟。在三维输运燃耗耦合程序MCORGS中增加了处理在线添加燃料与去除裂变产物的功能,实现了对液态燃料燃耗过程的模拟。增加6Li丰度和燃料初装量保持寿期初反应性不变,可以减缓寿期内反应性下降趋势。逐步增加包层内超铀元素装量,可以控制整个寿期内反应性基本恒定。聚变功率取20 MW,通过反应性控制,5年内包层能量放大倍数在160～180之间,氚增殖比在1.5～1.7之间,优于In-Zinerater基准设计方案。     

腐蚀和沉积对偏滤器靶板寿命的影响

讨论了偏滤器铍靶在聚变堆工况下的腐蚀问题，被物理溅射和辐射增强升华等过程腐蚀了的铍原子经电离后，部分将沉积到铍靶上，分析计算的结果表明，铍靶材被腐蚀后的自沉积将大大降低铍靶的净腐蚀率，在此基础，研究了一定浓度的硼杂质的沉积对铍靶的腐蚀率的影响，结果发现，铍靶前很低浓度的硼杂质即可大大降低了铍靶的腐蚀率，并且铍靶的腐蚀率随硼杂质浓度的增加而减少，这对聚变堆的稳态运行是有利的。     

加氚对D-3He系统聚变燃烧过程的影响

对加少量氚的D-³He聚变系统的点火燃烧过程进行了数值模拟研究,得到了有关的物理图象和一些主要计算结果。研究结果表明,加少量氚可以解决D-³He聚变系统的点火问题和加速其燃烧过程,从而提高燃耗。     

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.     

Compression and combustion of non-cryogenic targets with a solid thermonuclear fuel for inertial fusion

Variants of a target with a solid thermonuclear fuel in the form of deuterium-tritium hydrides of light metals for an inertial fusion have been proposed. The laser-pulse-induced compression of non-cryogenic targets, as well as ignition and combustion of such targets, has been examined. The numerical calculations show that, despite a decrease in the caloric content of the fuel and an increase in the energy losses on intrinsic radiation in the target containing deuterium-tritium hydrides of light metals as compared to the target containing deuterium-tritium ice, the non-cryogenic target can ensure the fusion gain sufficient for its use in the energy cycle of a thermonuclear power plant based on the inertial plasma confinement method.     

Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.     

Analysis of compression and burn of ion-beam inertial fusion targets including radiation transport

This paper presents implosion results of a new ion-beam inertial fusion target which has been designed for use in a reactor study, HIBALL-II. This target has been simulated using an updated version of the MEDUSA-KA code which includes radiation transport. The target contains 4 mg of DT fuel which is protected against radiative preheat by a high-Z, high-ρ lead radiation shield. The radiation shield is separated from the fuel by a reasonable thickness of lithium to improve the hydrodynamic stability. The target is driven by 10GeV Bi⁺⁺ ions and the peak power in the pulse is 500 TW. The total input energy is ～ 4.38 MJ and the gain is ～ 152.     

Optimal design of shock ignition target in order to stop generated hot electrons within the cold fuel

Generation of hot electrons (HEs) within ignitor pulse interaction with pre-compressed fuel is an important challenge in the shock ignition approach. Target optimization in order to prevent the destructive effects of HE is the main goal of the present work. In the first stage, the spectrum of electron energy generated during the interaction of ignitor pulse at different widths with the HiPER pre-compressed target has been estimated by applying particle simulation tool. Then, by changing the thickness of the cold fuel in the range of 185–225 μm, the corresponding areal densities are calculated using 1D hydrodynamic simulations. Finally, in order to assess the energy fusion yield, the iso-gain curves are obtained for different ignitor time windows as well as target thicknesses. Simulation results indicate that by decreasing the baseline, target thickness leads to a 17–70% increase in the fuel areal density. Subsequently, it has been demonstrated that by properly adjusting the parameters of ignitor pulse launch time and its width and employing a target with areal density high enough to stop the HEs, energy gain above 140 can be achieved. Optimal areas for shell thickness and ignitor time window are identified.     

基于现役装置的冲击点火可行性概念研究

以冲击点火物理特性的研究为基础, 分析冲击点火对高功率激光驱动器的物理需求, 然后从总体层面概括给出基于现役装置(神光III等间接驱动中心点火高功率激光装置) 研究冲击点火面临的关键技术问题. 研究表明, 基于现役装置的冲击点火主要面临两个层面的问题, 首先是非均匀光路排布下实现均匀辐照的工程层面问题, 其次是在现役装置上高效实现冲击点火激光脉冲的激光技术层面问题. 通过研究 分别对两个层面的问题提出相应的解决思路, 为后续研究奠定基础.     

Deuterium-tritium fuel impact ignition in the presence of degenerate plasma

The impact ignition model is proposed based on the collision of a deuterium-tritium (DT) layer accelerated to high velocities in a conical target. Simple mechanism, low cost, high coupling efficiency, and lack of the need for Petawatt laser pulses are the prominent advantages of this model. However, an increase in the productivity of this ignition mechanism is an important issue. In this regard, in this paper, the idea of impact ignition using the plasma degeneracy mechanism has been investigated. For this purpose, first, the ignition energy gain and stopping power of the DT beam in pure and impure fuels, by employing both degenerate and non-degenerate plasmas, have been examined numerically. Then, in order to assess the penetration depth and range of the incident beam, simulations have been carried out using a three-dimensional (3D) Monte Carlo code for two states of degenerate and non-degenerate pre-compressed pure fuel. The results imply that the state of degeneracy causes an increase by about 63% in the energy gain of impact ignition. In addition, the degeneracy condition leads to an approximate enhancement of 60% in the energy deposition of the pure fuel and about 67% for the impure fuel, with a mixed density ratio of 1.5%; therefore, the range and penetration depth decrease significantly in comparison to the non-degenerate one. This can be indicative of the increasing efficiency of impact ignition conditions in the presence of degenerate plasma. The results of the range for the pure fuel have also been confirmed by a 3D Monte Carlo simulation code.     

The influence of the state of a compressed thermonuclear substance on the combustion of inertial fusion targets under direct ignition by a short radiation pulse

One-dimensional numerical calculations were performed to study the dependence of conditions for initiating thermonuclear combustion and of the target gain of direct-ignition inertial fusion targets ignited by a short radiation pulse on the initial temperature of a preliminarily compressed fuel and the initial heat energy distribution between plasma electrons and ions in the ignition region (igniter). The igniter parameters at which an effective thermonuclear target explosion with a G ～ 10³ target gain occurred were shown to substantially depend on the initial temperature of the major fuel fraction and the initial heat energy distribution between igniter electrons and ions. The heat energy of the igniter passed a minimum as the size of the igniter decreased. The dependences of these minimum energies on the temperature of the major fuel fraction at various initial energy distributions between igniter electrons and ions were determined. An increase in the temperature of the major fuel fraction was shown to decrease the target gain.     

Ignition curves for deuterium/helium-3 fuel in spherical tokamak reactor

In this paper, ignition curve for deuterium /helium-3 fusion reaction is studied. Four fusion reactions are considered. Zero-dimensional model for the power balance equation has been used. The closed ignition curves for ρ=constant (ratio of particle to energy confinement time) have been derived. The results of our calculations show that ignited equilibria for deuterium /helium-3 fuel in a spherical tokamak is only possible for ρ= 5.5 and 6. Then, by using the energy confinement scaling and parameters of the spherical tokamak reactor, the plasma stability limits have been obtained in n_e, T plane and, to determine the thermal instability of plasma, the time-dependent transport equations have been solved.     

Diffusion filling with fuel gas of high-gain direct-drive cryogenic targets

To provide continuous operation of a reactor based on inertial confinement fusion (ICF), the thermonuclear burn region should be refilled with fuel with a frequency of 1 million targets per day. The first stage in the target production is diffusion filling of polymeric (CH) shells with fuel gas which is deuterium (D₂) or deuterium–tritium (DT) mixture. The results of simulation of filling reactor-scale CH-shells (Ø ~ 4 mm) to a pressure of ~1100 atm at 300 K in the mode with a constant pressure gradient are presented. Simple and two-layer shells of compact and porous polymers are considered. The problems of constructing an optimum DT-filling scheme avoiding CH-shell fracture due to tritium beta decay are discussed.