混合堆增殖乏燃料组件中子学特性初步研究

本文主要对聚变-裂变混合堆增殖乏燃料在压水堆组件中使用的可能性进行了初步研究。根据聚变 裂变混合堆增殖乏燃料的特点,给出了的聚变-裂变混合堆增殖乏燃料压水堆组件设计方案,分析组件的燃料温度系数、慢化剂温度系数等参数。结果表明：聚变 裂变混合堆乏燃料组件的特性与全铀组件的特性相似。在相同的易裂变同位素质量百分比情况下,本文给出的组件设计方案的功率不均匀系数更小。研究结果可为未来实现聚变 裂变混合堆和压水堆联合循环系统提供技术支持。     

核能源前景浅见

文章展望了裂变堆、纯聚变堆和聚变-裂变混合堆的前景,分析了混合堆的低聚变条件和很高的能量与燃料增殖能力等重大优点。认为作为由裂变能源过渡到纯聚变能源的桥梁,聚变-裂变混合堆应成为未来核能源的方向之一。     

聚变—裂变混合堆安全性初探

对聚变－裂变混合堆的安全性进行了初步分析和探讨。主要利用改进后的混合堆放射性程序ＦＤＫＲ对混合堆产生的核废物及放射性进行计算，并将结果与压水堆、高温气冷堆和液态金属冷却快中子增殖堆进行了比较。结果表明，混合堆与裂变动力堆相比有较好的安全性。     

聚变－裂变混合堆安全性初探

对聚变－裂变混合堆的安全性进行了初步分析和探讨．主要利用改进后的混合堆放射性程序ＦＤＫＲ对混合堆产生的核废物及放射性进行计算，并将结果与压水堆、高温气冷堆和液态金属冷却快中子增殖堆进行了比较。结果表明，混合堆与裂变动力堆相比有较好的安全性。     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源,采用天然铀为初装核燃料,并采用现有压水堆核电厂成熟的轻水慢化和冷却技术,设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析,研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示,现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料,都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求,但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

聚变-裂变混合堆的氚工艺问题

文章描述了聚变堆和聚变-裂变混合堆的氚工艺问题。根据聚变堆和聚变-裂变混合堆的特点讨论了对包层氚增殖材料的要求,列举了几种可作氚增殖的合理材料特性。给出了几种从包层提取氚和从废聚变燃料中回收氚的方法。最后对混合堆的氚安全及防护问题进行了讨论。     

我国核能发展中各种堆型的优化组合及混合堆的地位

为适应我国21世纪国民经济发展对能源的需求,寻找大力开发核能的途径,本文试图根据国际发展情况,我国核能资源条件,就高温汽冷堆、快堆(即裂变增殖堆)、聚变—裂变混合堆(即聚变增残堆)三种先进堆型的优化组合和我国核能发展进行一点探讨。     

聚变-裂变燃料工厂概念设计

报道了正在进行的一项聚变-裂变燃料工厂的概念设计计划,它的主要内容是进行参量系统的分析和进行一些关键性的实验与技术研究,并通过这些研究来探讨在中国建立聚变-裂变燃料工厂来支持PWR核电站的必要性和技术可行性,为下世纪初在中国建立一座实验性的聚变-裂变混合堆作可行性和方案性研究。已有的研究表明,在现有物理与技术基础之上,已有可能建成有意义的,以生产裂变燃料为主的聚变-裂变混合堆。     

PWR（U）乏燃料中超铀元素在混合堆快裂变包层内嬗变研究

从中子学角度对ＰＷＲ（Ｕ）乏燃料中的超铀元素（２３８Ｐｕ，２３９Ｐｕ，２４１Ｐｕ，２４１Ａｍ，２４３Ａｍ，２３７Ｎｐ，２４４Ｃｍ）在聚变－裂变混合堆快裂变包层内嬗变的可行性进了研究。利用一维中子输运和燃耗计算程序ＢＩＤＥＣＡＹ译不同燃料组分的四个快裂变包层进行分析计算。结果表明，在聚变－裂变混合堆快裂变包层内安全，高效地嬗变ＰＷＲ（Ｕ）乏燃料中的超铀元素是可能的。     

核聚变研究50年

分析了国内外核聚变研究成果现状和发展的趋势 ,对国民经济发展过程中的能源需求作了预测 ,对中国的聚变能源战略和历史机遇 (经济、技术体系、地位 )作了讨论 ,介绍了聚变 裂变混合堆并提出了发展聚变 裂变混合堆的总体设想、研究内容和预期目标。     

The case for the fusion hybrid

The use of nuclear fusion to produce fuel for nuclear fission power stations is discussed in the context of a crucial need for future energy options. The fusion hybrid is first considered as an element in the future of nuclear fission power to provide long term assurance of adequate fuel supplies for both breeder and convertor reactors. Generic differences in neutronic characteristics lead to a fuel production potential of fusion-fission hybrid systems which is significantly greater than that obtainable with fission systems alone. Furthermore, cost benefit studies show a variety of scenarios in which the hybrid offers sufficient potential to justify development costs ranging in the tens of billions of dollars. The hybrid is then considered as an element in the ultimate development of fusion electric power. The hybrid offers a near term application of fusion where experience with the requisite technologies can be derived as a vital step in mapping a credible route to eventual commercial feasibility of pure fusion systems. Finally, the criteria for assessment of future energy options are discussed with prime emphasis on the need for rational comparison of alternatives. This approach is contrasted with the dual standard too often used in judging the risks and benefits of nuclear power where, for example, rather minor radiological effects are highlighted while much larger exposures to radiation from medical x-rays, airplane travel, color television sets, etc., are ignored. It is concluded that the fusion hybrid deserves a prominent place among new energy resources but that early attention to insure an adequately informed public is a vital ingredient in assuring reasonable prospects of success.     

聚变-裂变混合堆外中子源效应

聚变-裂变混合堆(FFHR)作为聚变驱动次临界系统(FDS),具有良好的物理性能,能够实现产能、氚增殖、嬗变核废料等功能。采用COUPLE程序研究了水冷混合堆包层的铀水比和中子倍增剂对中子源效率的影响。结果表明:包层能谱越硬,外中子源效率越高;适当加入中子倍增剂Be可使外中子源效率增加。研究结果对进一步改进聚变-裂变混合堆的概念设计具有一定的指导意义。     

On the safety of conceptual fusion-fission hybrid reactors

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium.As a result of these studies, it appears that highly reliable and even redundant decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered.     

The fusion breeder

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U₃O₈). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-path-item for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices.     

Physico-economical premises for the nuclear design of fissile breeding devices

Physico-economical factors are the basis for guidelines in the nuclear design of emerging fissile breeding devices. In the present study 3 basic concepts are discussed: spallator, fusion-fission hybrid and the muon catalyzed fusion breeder. In all cases the expressions describing the income of a fissile breeder are given as functions of physical and technological system parameters. The dependence of the income on certain selected variables, others having been taken as parameters, is illustrated in a series of diagrams. An analysis of the obtained results indicates, among others, that: in all the above concepts high conversion ratios are desirable, thus making neutron slowing-down rather harmful; fast fissions in a spallator are advantageous; the plasma Q needs not be too high, but still should amount to ca. 5; the muon factory operating as a spallator is indispensable if the cold fusion (even with a rather optimistic efficiency) is to be economic.     

多功能聚变裂变混合实验堆FDS-MFX氦冷包层三维中子学初步设计与分析

FDS-MFX(Multi-Functional eXperimental fusion-fission hybrid reactor)是一个基于现实可行技术的多功能聚变裂变混合实验堆概念,分3个阶段相继开展实验研究,分别采用纯氚增殖包层、铀燃料包层和乏燃料包层.本文重点对其中铀燃料包层后期阶段中高浓缩铀模块的摆放方式...     

聚变-裂变混合堆设计研究

利用MCNP5和MONK9A程序对聚变驱动裂变混合堆进行了初步研究,在等离子体第1壁外侧依次包覆长方体形状的燃料组件和产氚组件,形成裂变堆芯包层和产氚区.对分别装载贫铀、天然铀、贫铀MOX和天然铀MOX等4种燃料的混合堆进行了研究分析,其中,后两种燃料在整个运行寿期内的功率放大倍数和氚增殖比满足设计要求.通过随燃耗变化的同位素含量分析,初步探讨了混合堆的铀-钚燃耗循环策略.     

Design of coated fuel particles for a hybrid fusion-fission system

Designs have been developed for coated ThO₂ fuel particles to be used in a hybrid fusion-fission system that could be operated without reprocessing. The fresh fertile fuel particle would first be cycled through the blanket of a fusion reactor to breed ²³³U, which would then be ‘burned’ in a thermal fission reactor. The depleted fuel would then be refreshed in a second pass through the fusion reactor, and the process above repeated as many times as feasible. Designs of coated particles for up to three cycles through the hybrid system of reactors have been developed. The outer structural layer for these particles is made from vapor-deposited silicon carbide, because of its remarkable dimensional stability under fast neutron irradiation, and an inner layer of porous pyrocarbon is used to accommodate the buildup of gaseous reaction products inside the particle. The production of gaseous emission products from the interaction of high-energy fusion neutrons with coating materials and with the oxygen in the kernel contributes significantly to pressure vessel stresses in these coatings, whereas gaseous fission products alone dominate in conventional thermal reactors. The most stringent design for the three-cycle particle is identical in fuel loading to the reference fertile particle for an HTGR, which would constitute an ideal hybrid partner for the fusion reactor. Consideration is also given to coated-particle designs for the containment of the bred tritium used to fuel the D-T fusion reactor.     

Reaction rates in blanket assemblies of a fusion-fission hybrid reactor

To validate neutronics calculation for the blanket design of fusion-fission hybrid reactor,experiments for measuring reaction rates inside two simulating assemblies are performed.Two benchmark assemblies were developed for the neutronics experiments.A D-T fusion neutron source is placed at the center of the setup.One of them consists of three layers of depleted uranium shells and two layers of polyethylene shells,and these shells are arranged alternatively.The ²³⁸U capture reaction rates are measured using depleted uranium foils and an HPGe gamma spectrometer.The fission reaction rates are measured using a fission chamber coated with depleted uranium.The other assembly consists of depleted uranium and LiH shells.The tritium production rates are measured using the lithium glass scintillation detector which is placed in the LiH region of the assembly.The measured reaction rates are compared with the calculated ones predicted using MCNP code,and C/E values are obtained.