小球形托卡马克嬗变堆堆芯参数分析

本文探讨作为聚变能中间应用的一种可行性 :利用低环径比球状托卡马克堆的高能聚变中子嬗变核电站的核废物。计算了堆芯等离子体物理参数、中心柱增益并进行参数学分析、几种可能的电流驱动方案比较和中心柱冷却方案设计及其计算。为了减轻偏滤器和第一壁材料的要求 ,我们有意选择较低堆芯物理参数 ,较低的中子壁负载运行。结果表明中心柱增益较低。建议从理论上探索研究由球状托卡马克等离子体顺磁性 ,建立一个无力球马克极向电流壳为中心区提供大部分或全部Bt,旨在建立无中心柱的低径比球状托卡马堆。如果可行 ,它的性能将会大大改善     

聚变堆第一壁TiC涂层材料研究

本文介绍了第一壁ＴｉＣ涂层材料的化学气相沉积（ＣＶＤ）工艺；研究了涂层工艺条件对涂层微观结构、沉积速率和基体对涂层生长的影响；给出了涂层厚度与涂层工艺条件之间的经验关系式。讨论了ＴｉＣ／石墨、ＴｉＣ／ＭＯ、ＴｉＣ／３１６ＬＳＳ涂层材料的电子束热冲击损伤机理和热疲劳损伤机理，并对ＴｉＣ／石墨、ＴｉＣ／Ｍｏ、ＴｉＣ／３１６ＬＳＳ用做现行Ｔｏｋａｍａｋ第一壁材料的可行性进行了分析。     

低环径比设计概念——一种可接受的核聚变动力系统

低环径比（ＬＡＲ）概念作为一种聚变动力堆设计方案,它的许多关键问题已用总体设计程序进行了评估计算。Ａ＝１．４,κ＝３,βＴ＝６２％和自持电流为８７％的平衡设计的物理基础取自过去已得到的结果。计算表明,若引入氪（Ｋｒ）以便增加等离子体芯部辐射,则可将偏滤器上的热流部分地转移到第一壁上,但此时芯部的聚变功率也有所减少。设计中采用了积极的工艺措施,将关键部件的参数推向设计极限,为了减小堆尺寸,对铜合金的环向磁场（ＴＦ）线圈中心柱未施屏蔽。本文提出了一项能承受高的热负荷和中子壁负载的堆设计,它在第一壁和包层中用氦作载热剂、钒合金作结构材料、锂作氚增殖剂。若采用上述材料组合和设计,一个电功率为１９６６ＭＷｅ的聚变动力堆的尺寸,可减小到大半径为３．０５ｍ。设铜合金和钒合金的辐照寿命为１５ＭＷｙ／ｍ２,考虑到在堆的３０年寿期内,须更换铜合金中心柱和外侧第一壁以及包层,估计电价为５．１９美分／千瓦时,而总的直接投资为４８．８亿美元。这说明采用本ＬＡＲ设计,聚变电厂的造价（２４８２美元／千瓦电）开始接近可接受的价格水平,为了获取经济的聚变动力,采用ＬＡＲ发展途径可能是明智之举。     

在托卡马克第一壁上中子流量及壁负荷的分布

利用 MCNP 方法编制了程序 DNWLMC 并用 DNWLMC 计算了托卡马克第一壁上中子壁负荷的分布和中子流量对中子入射角余弦的分布。计算结果表明,中子壁负荷的分布和中子流量对入射角余弦的分布是复杂的,第一壁的位形、第一壁和等离子体的相对位置对此有重要影响。一般情况下,中子壁负荷分布极不均匀;中子流量对入射角余弦的分布表现为以大角度入射的中子很少,大部分中子垂直入射或是接近垂直入射。     

低环径比聚变托卡马克液态金属包铜中心导体柱

低环径比(LAR)聚变托卡马克可作为一种紧凑中子源,为聚变早期利用提供了有效途径。本文针对其关键部件中心导体提出新颖液态金属(LM)包钢中心导体往设计结构,其结构由既可增殖氚又可降低导体辐照损伤的防护层和负载巨大电流(>10MA)的中心导体区组成。与常规中心导体柱(CCP)相比,其优势在于:防护区减小了铜导体的辐照损伤;提高了中子利用率及氚增殖率;优化结构的铜导体,保证了较低的欧姆损耗。     

聚变堆螺旋流道液态第一壁流动稳定性分析

聚变堆包层第一壁材料所面临高能粒子辐照、电磁辐射、高热负荷、复杂的机械负荷和相应的物理化学腐蚀制约其服役性能和使用寿命,是聚变能发展的瓶颈问题。液态第一壁由于液态工质自身的特点可以承受更高的热负载、中子壁负载以及更高的出口温度,且由于液态工质的不断更新不存在中子辐照损伤问题,在未来聚变堆应用中很具有吸引力。但由于液态金属在聚变堆强磁场作用下流动形成磁流体(Magnetohydrodynamic MHD)效应,维持液态第一壁在复杂的几何结构和苛刻的工作条件的稳定流动性是现有液态壁研究的难点问题。本文针对自由表面液态金属流动时产生的MHD特性,提出了螺旋流道液态壁流动方案,通过在真空室背壁上设置沿磁场方向的螺旋型流道,使流道内液态金属沿磁场运动,进而减少切割磁场产生的MHD效应。并参考典型聚变堆FDS-Ⅱ,建立了外包层三维模型与真实磁场位型,对方案进行MHD分析与优化,分析结果表明该方案可以在真空室表面形成完整、稳定的液态金属包裹,验证了该方案在磁场作用下液态第一壁流动稳定性与初步可行性。     

聚变堆实验包层模块第一壁温度场及热结构分析

铍相对于众多聚变反应堆的第一壁护甲材料,有着许多优点,这些优点使它和钨及碳基材料一起被选作国际热核聚变实验堆（ITER）第一壁的候选防护材料。对中国氦冷固态增殖剂实验包层模块（CHHCSBTBM）第一壁进行多场耦合模拟分析结果表明,使用表面热负荷模拟分析时,未考虑中子负载情况下,模拟分析结果与其它结果有较大出入,故使用表面热负荷模拟分析时必须考虑中子负载情况。而对第一壁热结构分析表明,铍保护板的应力超过了其许用应力,可以寻找其它铍合金或第一壁护甲材料以满足第一壁护甲材料热结构应力要求。     

一个研究聚变中子辐照损伤的蒙特卡罗程序

文章描述计算机模拟聚变中子辐照损伤的蒙特卡罗程序。该程序适用于含轻，重核素的非晶态复合材料及能量低于１５ＭｅＶ的中子和各种能量的离子。文中给出辐照后靶原子Ｆｅ，Ｃｒ，Ｎｉ，Ｍｏ，Ｗ，Ｓｉ，Ｃ的ＤＰＡ截面，ＰＫＡ能谱和发射角分布的计算结果，并给出作为混合堆第一壁材料的３１６ｓｓ，钨，石墨，碳化硅化移损伤率的气体产生率。     

LiPb自冷混合堆包层中子学研究

研究了ＬｉＰｂ自冷托卡马克混合堆包层的中子学性能；第一壁材料和厚度对中子学性能的影响；Ｐｂ和Ｂｅ的中子增益性能以及包层中功率密度和２３９Ｐｕ的分布，并对中子学性能进行了优化。当聚变功率为２００ＭＷ，运行因子为０．３时，除氚自给外，每年可生产２３９Ｐｕ１３０ｋｇ。     

cosRMC复杂曲面开发及在聚变中子学的应用

中国聚变工程实验堆（CFETR）是我国自主设计和研制的重大科学工程，CFETR旨在与ITER相衔接和补充，为研制DEMO级别聚变堆电站提供必要的技术。蒙特卡罗方法在聚变中子学与屏蔽设计等方面具有重要作用。本文基于自主化蒙特卡罗程序cosRMC，研究了蒙特卡罗复杂曲面建模的数学模型和计算方法，开发了复杂曲面建模功能，并通过PPCS（power plant conceptual study）模型验证了该功能实现的正确性。然后构建了CFETR的三维精细化模型，并利用该模型对CFETR包层设计中的关键中子学参数进行计算分析。结果表明，cosRMC对中子学参数氚增殖比、中子壁载荷和核热沉积的计算结果与MCNP的计算值吻合良好，相对偏差均小于5%，满足工程设计需求。研究证明了cosRMC应用于聚变堆包层中子学分析的正确性和有效性。CFETR中子学参数的计算分析，也为其设计和优化提供了参考。     

Design modification and optimization of the ITER cooling water system

ITER (Latin for “the way”), the largest fusion experimental reactor in the world, is designed to demonstrate the technological feasibility of nuclear fusion energy conversion, at plant scale, from high temperature deuterium-tritium plasma using the Tokamak magnetic confinement arrangement.ITER will have a large vacuum vessel that hosts the plasma facing components. These components include the blanket and the divertor that will operate at temperatures, heat loads, and neutron flux higher than those reached in a nuclear fission power plant reactor.One of the main critical issues of the ITER reactor is the design of the cooling water system to transfer the heat generated in the plasma to the in-vessel components and the heat loads from the auxiliary systems to the environment.This paper describes the current ITER cooling water system and recent design modifications and optimizations.     

Radiation Damage and Tritium Breeding Study in a Fusion Reactor Using a Liquid Wall of Various Thorium Molten Salts

A new magnetic fusion reactor design, called APEX uses a liquid wall between fusion plasma and solid first wall to reach high neutron wall loads and eliminate the replacement of the first wall structure during the reactor’s operation due to the radiation damage. In this paper, radiation damage behavior of the inboard and outboard first walls made of a ferritic steel, 9Cr-2WVTa, in the APEX blanket using various thorium molten salts, 75% LiF-25% ThF₄, 75% LiF-24% ThF₄-1% ²³³UF₄ and 75% LiF-23% ThF₄-2% ²³³UF₄ was investigated. Furthermore, tritium breeding potential of these salts in such a blanket was also examined. Computations were carried out using the code Scale 4.3 by solving Boltzmann neutron transport equation. Numerical results brought out that only the liquid wall containing the molten salt, 75% LiF-23% ThF₄-2% ²³³UF₄ and having a thickness of ≥38 cm would be suitable to be used in the APEX reactor with respect to radiation damage criteria for the first wall structures and tritium self-sufficiency for the (DT) fusion driver.     

The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.     

Monte Carlo calculation of radiation damage in first wall of an experimental hybrid reactor

This work was focused on the neutronic calculation of the nuclear parameters (neutron spectrum, displacement per atom (DPA), gas production, tritium breeding ratio (TBR), nuclear heating) for structural materials in the first wall (FW) and fuel clad (made of ferritic/martensitic steels, vanadium alloy, silicon carbide, copper alloy, and stainless steel) of an experimental hybrid reactor using the most current Monte Carlo Neutron-Particle Transport code MCNP5 1.4. Neutronic calculations were performed using a (DT) fusion driver hybrid reactor under a neutron wall loud of 2.25 MW/m² by full reactor power for one year. Obtained results were compared with three different data libraries (ENDF/B-V, ENDF/B-VI and CLAW-IV). TBR values in the reactor blanket for all investigated materials became greater than the minimum requirement (TBR > 1.05). Nuclear parameters like DPA, He-production and nuclear heating were considered as radiation damage limits for structural materials, copper alloy (Cu0.5Cr0.3Zr) showed better performance than all investigated materials.     

Transmutation-incineration potential of transuraniums discharged from PWR-UO2 spent fuel in modified PROMETHEUS fusion reactor

This study presents the potential of the burning and/or transmutation (B/T) of transuraniums (TRUs), discharged from the pressured water reactor PWR-UO₂ spent fuel, in the modified PROMETHEUS-H fusion reactor. Two different design shapes (Models A and B) were considered. The transmutation zone (TZ), which contains the mixture of TRU nuclides (10%), was located in the modified blankets. The volume fraction of Pu in the mixture is raised from 0 to 40% stepped by 10% to determine its effect on the B/T. The fuel spheres were cladded with SiC (5%) and cooled with high-pressured helium gas (85%) for nuclear heat transfer. The calculations were performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a neutron wall load (P) of 4.7 MW/m². The results bring out that: (1) the Model B transmutes the TRUs more rapidly than the Model A; (2) the effective half-lives decrease about 20 and 40% with the increase of Pu fraction in the cases of Models A and B, respectively; (3) the M values are quite high with respect to the M value of the original PROMETHEUS fusion reactor; (4) the blankets can produce substantial electricity in situ.     

Fusion neutron induced radioactive emissions from surfaces

Neutron induced direct nuclear recoil sputtering ratios have been measured for a variety of 14.8 MeV (d, t) neutron induced reactions in Nb, Mo, V and 316 SS. Absolute recoil sputtering ratios for forward and backward sputtering are reported. Forward sputtering ratios are typically in the range of 10^?9–10^?7 recoil atoms per (d, t) neutron while backward sputtering ratios are usually several orders of magnitude lower. Some of the implications of radioactive particle ejection in the first wall region of fusion reactors are discussed. It is shown that radioactivity ejected into the fusion reactor coolant channels by direct nuclear recoil and by lattice dynamic neutron sputtering, may have a significant effect on the design and maintenance of fusion reactors.     

CFETR第一壁及赤道面外包层中子辐照损伤初步分析

中国聚变工程实验堆(Chinese Fusion Engineering Testing Reactor,CFETR)的包层和偏滤器第一壁面向堆芯等离子体,第一壁辐照损伤分析对于托克马克安全运行至关重要。赤道面外包层较其它包层距离堆芯等离子体中心更近,其结构材料承受中子辐照大。因此,进行中子辐照损伤评估十分必要。基于此目的,采用计算机辅助设计(Computer Aided Design,CAD)模型和蒙特卡罗中子学建模转换接口Mc CAD完成中子学建模,并用蒙特卡罗方法的粒子输运程序计算第一壁和氦冷固态外包层结构材料辐照损伤。此外,对比了铍和钨作为面向等离子体材料两种情况下第一壁的受损情况。计算结果表明,氦冷固态包层模型下结构材料可以满足CFETR一期的运行要求。     

Irradiation effects on CTR first wall materials

The selection of the controlled thermonuclear reactor (CTR) first wall material (refractory metallic alloy or austenitic stainless steel) will be a compromise based on a number of important nuclear, physical, thermal and mechanical properties, and safety, environmental, technological and economic factors. The correlations between helium production, irradiation, and helium embrittlement in the first wall depend mainly on the neutron spectrum, neutron fluence, irradiation temperature, wall material, reactor operating conditions, etc. The suppression of irradiation swelling by alloying elements (C, Si, Mo, and P) is effective and beneficial. The extent of change in mechanical properties of the first wall material due to neutron irradiation, thermal instabilities (thermal shock, thermal fatigue, crack initiation and propagation), corrosion effect, etc. remains undefined or not completely understood. A fabricated CTR first wall material must meet the requirements of design safety, weldment reliability and good operation performance.     

Recent progress in research on tungsten materials for nuclear fusion applications in Europe

The current magnetic confinement nuclear fusion power reactor concepts going beyond ITER are based on assumptions about the availability of materials with extreme mechanical, heat, and neutron load capacity. In Europe, the development of such structural and armour materials together with the necessary production, machining, and fabrication technologies is pursued within the EFDA long-term fusion materials programme. This paper reviews the progress of work within the programme in the area of tungsten and tungsten alloys. Results, conclusions, and future projections are summarized for each of the programme’s main subtopics, which are: (1) fabrication, (2) structural W materials, (3) W armour materials, and (4) materials science and modelling. It gives a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments.