A model for tritium transport in fusion reactor components: The FUS-TPC code

Hydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of atomic hydrogen in materials in which hydrogen and its isotopes are present. In this work the problem of tritium transport from lead–lithium breeder through different heat transfer surfaces to the environment has been studied and analyzed by means of a computational code. The code (FUS-TPC) is a new fusion-devoted version of the fast-fission one called Sodium-Cooled Fast Reactor Tritium Permeation Code (SFR-TPC). The main features of the model inside the code are described. A simulation, using the code, was performed by adopting the configuration of the European configuration of the Helium Cooled Lead Lithium (HCLL) blanket for DEMO.     

A preliminary conceptual design study for Korean fusion DEMO reactor

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb₃Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

Surface bombardment rates for mirror fusion reactor designs

Arrival rates of D⁰, T⁰, D⁺, T⁺, He⁺⁺, neutrons, and photons are given for FERF (Fusion Engineering Research Facility, a mirror confinement reactor dedicated to materials research and to component testing), a hybrid fusion-fission reactor designed primarily to produce fissile fuel, and a D-T power reactor design. For comparison a next-generation confinement-mirror experiment called MX is included. The surfaces of interest are the first wall, the end wall, the direct converter and the injector.     

Effect of outside tritium source on tritium balance of a D-T fusion reactor

Attainable tritium breeding ration in the blanket system must be larger than the required breeding ratio when no effective tritium resources from outside are expected. It is revealed recently that a considerable amount of tritium can be trapped to the re-deposition layer of the first wall materials and that the time constant of this phenomenon is rather long. Then, the tritium breeding ratio around 1.1 is required in the blanket system when 3 years is claimed for the tritium doubling time to prepare tritium for the initial inventory of a next reactor. Construction of an outside tritium supply is one of the possible ways to compensate the lack of tritium because it is generally considered that the attainable tritium breeding ratio in the solid breeder system is around 1.05. It is reported recently that a high-temperature gas-cooled reactor can produce 10 kg of tritium per year. The preferable amount of tritium production rate of the outer tritium supply is discussed in this study from the viewpoint of tritium balance in a D-T power reactor.     

Tritium handling, breeding, and containment in two conceptual fusion reactor designs: UWMAK-II and UWMAK-III

Tritium is an essential component of near-term controlled thermonuclear reactor systems. Since tritium is not likely to be available on a large scale at a modest cost, fusion reactor designs must incorporate blanket systems which will be capable of breeding tritium. Because of the radiological activity and capability of assimilation into living tissues, tritium release to the environment must be strictly controlled. The University of Wisconsin has completed three conceptual designs of fusion reactors, UWMAK-I, UWMAK-II, and UWMAK-III. This report discusses the tritium systems for UWMAK-II, a reactor design with a helium cooled solid breeder blanket, and UWMAK-III, a reactor design with a high-temperature liquid breeder blanket. Tritium systems for fueling and recycling, breeding and recovery, and plant containment and control are discussed.     

聚变堆中性束注入系统的电偏转器概念设计

中性束注入是磁约束核聚变研究重要的辅助加热和电流驱动手段.由于负离子源中性束注入系统束能量高、束斑大,电偏转已经成为剩余离子剥离的首选方案,其剩余离子剥离设备被称为电偏转器.电偏转器作为实现束流中性化的核心设备,其性能决定了中性束注入系统的工作效率.本文提出了聚变堆主机关键系统研究中负离子源中性束注入系统电偏转器的概念...     

Chemical thermodynamics of fusion reactor breeding materials and their interaction with tritium

Liquid lithium, lithium alloys (solid and liquid) and ceramic lithium compounds are candidate breeding materials for (D,T) fusion reactors.Besides their tritium breeding capability, which results from neutron capture, their thermochemical properties and their interaction with tritium are of particular interest. A good knowledge of the physical and chemical properties of liquid lithium exists; and the systems Li-LiH, Li-LiD and Li-LiT have been studied in great detail. For dilute solutions of D₂ in liquid lithium, Sieverts' law was found to be valid down to an atom fraction of $\text{x}{}_{\mathrm{D}}\text{= 10}{}^{-6}$; in the vapor, lithium polymers up to Li₄ and lithium deuterides are found.In the system liquid Li-Pb, the solubility of D₂ was measured as a function of temperature and alloy composition, and correlated with the activities of the constituent metals. The solubility of D₂ was found to obey Sieverts' law at low concentrations, and is many orders of magnitude smaller than that in liquid lithium. This holds also for solid “Li₇ Pb₂”.Vaporization studies yielded data on the thermal stability of the oxides: Li₂0, γ-LiAlO₂, β-Li_sAlO₄, LiAl₅O₈, Li₂ZrO₃, Li₄ZrO₄, Li₈ZrO₆, Li₂SiO₃ and Li₄SiO₄. Tritium diffusivity was studied in Li₂O, γ-LiAlO₂, β-Li₅AlO₄ and Li₄SiO₄. A large number of gaseous lithides were detected during these studies.     

Physical sputtering model for fusion reactor first-wall materials

A model has been developed to provide energy-dependent physical sputter yields for various plasma particles (ions and neutrals) incident on candidate first-wall materials. The physical sputter yield is expressed in terms of the atomic and mass numbers of the projectile and target atoms, the surface binding energy of the targets and the energy of the incident particle. The general shapes of the yield curves are based on theoretical models, whereas the magnitudes of the yields are derived primarily from experimental data. The model applies to both high- and low-Z incident particles bombarding high- and low-Z wall materials. Although the model was developed for metal first-wall materials, it has been extended, with minor modifications, to predict physical sputter yields of several stable compound wall materials. A comparison of predicted yields with available experimental data has been made for a number of candidate first-wall materials.     

Development of a hydrogen permeation sensor for future tritium applications

Tritium monitoring in lithium–lead eutectic is of great importance for the performance of liquid blankets in fusion reactors. In addition, tritium measurements will be required in order to proof tritium self-sufficiency in liquid metal breeding systems. On-line hydrogen (isotopes) sensors must be design and tested in order to accomplish these goals.In this work, an experimental set up was designed in order to test the permeation hydrogen sensors at 500 °C. This experimental set-up allowed working with controlled environments (different hydrogen partial pressures) and the temperature was measured using a thermocouple connected to a temperature controller that regulated an electrical heater.In a first set of experiments, a hydrogen sensor was constructed using an α-iron capsule as an active hydrogen area. The sensor was mounted and tested in the experimental set up. In a second set of experiments the α-iron capsule was replaced by a welded thin palladium disk in order to minimize the death volume. The experiments performed using both membranes (α-iron and palladium) showed that the operation of the sensors in the equilibrium mode required at least several hours to reach the hydrogen equilibrium pressure.     

The formation of tritium permeation barriers by CVD

The effectiveness as permeation barriers of the following CVD coatings have been investigated: TiC (1 to 2 μm in thickness); a bi-layer of TiN on TiC (3 μm total thickness) and CVD A1₂O₃ on a TiN/TiC bi-layer. The substrate materials were TZM (a Mo alloy) and 316L stainless steel in the form of discs of diameter 48 mm and thickness 0.1 or 1 mm. Permeation measurements were performed in the temperature range 515–742 K using deuterium at pressures in the range 1–50 kPa. CVD layers were shown to form reasonably effective permeation barriers. At a temperature of 673 K TiC is around 6000 times less permeable to deuterium than 316L stainless steel.     

新型反应堆可视化设计方法研究

以核聚变反应堆为研究对象,提出了一种参数化的零部件设计为先导,以可变异的零部件组装逻辑为手段的新型的可视设计方法,所有的设计对象采用实体造型技术,可对系统的设计结果进行三维渲染和动态仿真。设计的对象包括偏滤器、中心柱、包层、纵向场、极向场等,以偏滤器为例,部件组成又包括轨道、支体、外靶板、内靶板、能量沉积板、圆顶和翼。所介绍的设计系统为核聚变装置的灵活设计提供了一种有效的手段,同时也将促进核聚变装     

SiC薄膜的阻氚性能研究

在不锈钢基体表面用离子束混合技术沉积SiC薄膜,然后用能量为5 keV的H+对其辐照直至剂量达到1×1018/cm2,再用二次离子质谱分析(SIMS)分析H+在SiC薄膜中深度分布和正离子谱,研究薄膜的阻氢性能和阻氢机理;最后采用渗透实验对涂覆在不锈钢基体表面的SiC材料的氚渗透系数进行测试,对其阻氚性能进行验证.结果表明,在不锈钢基材表面涂覆的SiC薄膜具有良好的阻氢性能,可将氚的渗透率降低4个数量级以上,SiC薄膜的阻氢是由于氢与薄膜中的硅、碳悬挂键反应形成诸如C-H、C-H2、C-H3、Si-H、Si-H2和Si-H3引起的.     

ITER中国液态锂铅实验包层模块氚渗透分析

对ITER中国液态锂铅实验包层模块的氚渗透途径进行了初步分析,并建立了氚渗透模型;在确保环境安全的前提下,通过计算LiPb中的氚分压分析了氚渗透量及氚总量的分配情况;在此基础上通过改变进入氚提取系统中LiPb比例(F)和涂层氚渗透减少因子(TPRF)对氚提取及渗透的影响做了灵敏性分析.     

Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor

The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a near-term production reactor. Four tokamak reactor concepts (TORFA/FED-R, AFTR/ZEPHYR, Riggatron, and Superconducting Coil) of approximately 500-MW fusion power are compared with regard to their demands on plasma performance, required fusion technology development, and blanket configuration characteristics. Because of its relatively moderate requirements on fusion plasma physics and technology development, as well as its superior configuration of production blankets, the TORFA/FED-R type of reactor operating with a fusion power gain of about 3 is found to be the most suitable tokamak candidate for implementation as a near-term production reactor.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing from its 1982 version (except for Tables II and III and Fig. 1), explaining the fact that some of the material is dated.