加速增产核燃料的天然安全“核热泉”快中子增殖堆

为保证21世纪中国经济的持续稳定地高速增长,必须充分发挥核能的巨大潜力,使之配合其他可再生能源同步增长,及早大规模替代煤炭等化石能源。由于目前国内大量兴建的核电站以压水堆为主,需要消费大量天然铀资源,倚靠廉价铀供应难于维持长期增长,必须依靠快中子增殖生产人造裂变燃料——钚,才能摆脱天然铀原料短缺的束缚。然而,传统的快中子增殖堆的核燃料增产速度较慢,难于配合中国核电的高速增长。本文介绍一种先进快中子增殖堆(AFBR)方案,其中利用在线连续换料的空心球形燃料元件,依靠载热剂的出入口之间的温度差实现满功率自然循环,可以成倍地提高燃料比功率与核燃料增殖速度。本快中子增殖堆改进了俄罗斯称为"天然安全"的BREST铅冷快堆设计方案,成为无须人为控制的"核热泉",它能在不设置加压泵及高位铅池的情况下,自动按外部负荷需要供应必要的热量,完全依靠自然循环将全部裂变热能及停堆后堆芯余热散出,不至对环境产生放射性污染。     

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.     

聚变堆液态金属锂铅包层多功能涂层研发

液态金属锂铅包层是目前国际上聚变堆包层设计研究的主要方案之一,结构材料表面制备涂层是降低锂铅包层中的氚渗透率、液态锂铅腐蚀及磁流体动力学(MHD)效应的重要技术之一.本文主要从涂层材料及其制备工艺两个方面重点介绍了国内外在液态锂铅包层涂层材料研发方面的进展概况,并对涂层技术发展进行了展望,最后提出了中国发展液态锂铅包层涂层的规划建议.     

Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.     

Design principles for efficient muon production in muon-catalyzed fusion reactors

The two major schemes for a fixed-target muon production system in muon-catalyzed fusion reactors are analyzed and compared using Monte Carlo simulation techniques. Starting with a careful optimization of the pion production target we next consider the complete system where pion conversion losses and muon losses in the target and the pressure vessel are taken into account. A simple but realistic design for the pion-muon converter is introduced. Problems and inefficiencies are identified to provide a basis for future inventions.     

Neutron parameters which can be reached in fast uranium blankets of hybrid fusion reactors

Conclusions The above considerations have shown that when the real requirements to blanket design are taken into account, a production of 1.4–1.8 tons of plutonium nuclei per fusion event can be expected in a blanket in which metallic uranium fuel is employed. Accordingly, in a reactor with a fusion power of 500 MW, 2.5–3 tons of plutonium can be produced per calendar year; when the total thermal power of the reactor is 2.5–3 GW at the beginning of the operational period, the breeding amounts to 1 kg/(MW·yr).Translated from Atomnaya Énergiya, Vol. 57, No. 1, pp. 36–41, July, 1984.     

Optimized fissile and fusile breeding in a laser-fusion fissile-enrichment flux trap blanket

Optimization of fissile and fusile production in the SOLASE-H laser-fusion fissile-enrichment fuel-factory blanket is carried out. The objective is maximizing fissile breeding with the constraints of maintaining self-sufficiency in tritium production, and realistically accounting in the modeling for structural and coolant compositions and configurations imposed by the thermal-hydraulic and mechanical designs. The effect of radial and axial blanket zone thicknesses on fusile and fissile breeding is studied using a procedure which modifies the zones' effective optical thicknesses, rather than the actual three-dimensional geometrical configurations. A tritium yield per source neutron of 1.08 and a Th (n, ) reaction yield per source neutron of 0.43 can be obtained in such a concept, where ThO₂ Zircaloy-clad fuel assemblies for light water reactors (LWRs) are enriched in the²³³U isotope by irradiating them in a lead flux trap. This corresponds to 0.77 kg/[MW(th)-year] of fissile fuel production, and 1.94 years of irradiation in the fusion reactor to attain an average 3 w/o fissile enrichment in the fuel assemblies. For a once-through LWR cycle, a support ratio of 2–3 is estimated. However, with fuel recycling, more attractive support ratios of 4–6 may be attainable for a conversion ratio of 0.55, and of 5–8 for a conversion ratio of 0.70. These estimates are lower than those reported, around 20, for related designs.     

Neutron multiplier alternatives for fusion reactor blankets

A proposal is made to replace the neutron multiplier in fusion reactor blankets by an efficient moderator (⁷LiH or ⁷LiD). The advantageous effect of the intensified neutron-energy degradation is due to the $\text{1}\text{v}$ character of the main tritium-producing reaction. The slowing-down medium is designed to be the source of moderated neutrons for the surrounding Li region where the most of the tritium is to be produced. The surplus tritium produced remains stored in the moderator zone. Some preliminary calculations illustrating the above concept were carried out, and the neutron flux and tritium production distributions are presented. Indications regarding further studies are also suggested.     

星座型裂变燃料核反应堆的物理构想

从分析钍在持续中子辐照过程中各代子体含量的演变出发,着重研究有多代子体均达到各自的饱和值时的情况和所具有的特性,提出星座型裂变物质核反应堆的物理构想,并就此堆的特性和应用前景作了简单阐述和讨论。     

First wall thermal hydraulic models for fusion blankets

Subject to normal and off-normal reactor conditions, thermal hydraulic models of first walls, e.g., a thermal mass barrier, a tubular shield, and a radiating liner, are reviewed. Under normal operation the plasma behaves as expected in a predicted way for transient and steady-state conditions. The most severe thermal loading on the first wall occurs when the plasma becomes unstable and dumps its energy on the wall in a very short period of time (milliseconds). Depending on the plasma dump time and the area over which the energy is deposited melting of the first wall surface may result, and if the temperature is high enough, vaporization.     

Reactor physics ideas for large scale utilization of thorium in gas cooled reactors

The physics principles for maximizing the fertile to fissile conversion were used in developing reactor concepts for large scale utilization of thorium in thermal and fast reactors (Jagannathan & Pal, 2006; Jagannathan et al., 2008). It is recognized that these principles are very well suited for ‘He’ gas cooled reactors with graphite moderator since both helium gas coolant and the graphite moderator have low neutron absorption characteristics and thus gives better neutron economy. In this paper, these ideas are applied to the High Temperature Test Reactor (HTTR) core of Japan to assess its advantage over the present day gas cooled reactors. HTTR is helium cooled and graphite moderated system. Significant amount of thorium has been loaded in the HTTR core with some minimal changes in the existing core design. The modified design is called HTTR-M core.In the HTTR-M core, the fuel is changed from enriched UO₂ fuel to Pu in ThO₂ fuel. The locations of boron type burnable poison rods within each fuel assembly of HTTR are replaced by one cycle irradiated thoria rods. Also, the B₄C type control assembly around the HTTR core is replaced by fresh seedless thorium assembly. The fertile thoria assembly are scattered uniformly in the HTTR-M core. The equilibrium core of HTTR-M shows very small burnup reactivity swing. The core excess reactivity is ∼18 mk at BOC and reduces to 1 mk at 660 days. It is interesting to note that this small reactivity change is intrinsically achieved by the choice of seed and fertile dimensions and their contents without the use of burnable poison rods or mechanical control rods which are used in HTTR core. The burnup reactivity swing in the latter after using burnable poison is ∼100 mk. The fissile seed inventory ratio (FIR) in a fuel cycle is 0.90 as compared with 0.717 of HTTR core. Since ²³³U is a better fissile nuclide with highest ‘η’ value in thermal range, the above conversion ratio can be regarded as quite good.     

Plutonium fuel and fuel elements for power reactors

Conclusions The use of plutonium in the fuel cycle during complex utilization of thermal and fast reactors in nuclear energetics permits solving the problem of ensuring nuclear fuel for a long period. Oxide uranium-plutonium fuel facilitates the development of technology of fast reactors and so far it is considered as the basic type of fuel. At the same time, oxide fuel cannot ensure the required rate of plutonium accumulation, in view of which the investigations of more efficient fuel and constructional materials become a pressing problem. The use of uranium-plutonium oxide fuel in thermal reactors requires improvements in the construction of fuel elements and organization of large-scale completely automatic production.Translated from Atomnaya Énergiya, Vol. 43, No. 5, pp. 412–417, November, 1977. Editors' Remarks. For the completeness of the discussion of the problem it is, of course, necessary to consider the possibility of using plutonium in fast and thermal reactors as done by the authors. However, it should be kept in mind that by its nuclear-physical parameters plutonium as a nuclear fuel is more suitable for use in fast reactors than in thermal reactors. The use of plutonium in thermal reactors can reduce the demands of natural uranium for the development of nuclear power in all by 10–15%, whereas its use in fast reactors reduces the demand for uranium by a factor of 10.All this indicates the feasibility of using plutonium only in fast reactors even if its accumulation is required over a certain period.     

New concept of designing Pu and MA containing fuel for fast reactors

New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO₂ powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO₂ powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.     

Compound cryopump for fusion reactors

We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium “ash” is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15–22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ～30 K. In a separate regeneration step, deuterium and tritium are released at ～110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly.The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce the required tritium reserve by almost 90%.We used a well-established free Direct Simulation Monte Carlo (DSMC) code, DS2V. At very high upstream densities (～10²⁰ molecules/m³ and above) the flow into the pump is choked. Enlarging the aperture is the only way to increase the pumping speed at high densities. Ninety percent of the deuterium and tritium is successfully trapped at 15 K (assuming that the sticking coefficient is 80–100% on the 15–22 K surface). On the other hand, the remaining 10% still exceeds the small amount of helium in the gas input.     

Development of the breeding blanket and shield model for the fusion power reactors system SYCOMORE

SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory.Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper.For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available.Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE.In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006–2007 studies) is also presented.     

聚变—裂变混合堆及其在我国核能发展中的作用

本文概要介绍聚变和聚变-裂变混合堆基本原理及其作用。聚变-裂变混合堆可以为压水堆或快堆提供充足的核燃料。它和压水堆或快堆组成的系统具有经济可行性。在解决我国核能发展中燃料短缺问题和促进纯聚变能源的发展方面可望发挥重要的作用。     

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...