Proposal of a Deuterium-Deuterium Fusion Reactor Intended for a Large Power Plant

This article looks for the necessary conditions to use Deuterium-Deuterium (D-D) fusion for a large power plant. At the moment, for nearly all the projects (JET, ITER…) only the Deuterium-Tritium (D-T) fuel is considered for a power plant. However, as shown in this article, even if a D-D reactor would be necessarily much bigger than a D-T reactor due to the much weaker fusion reactivity of the D-D fusion compared to the D-T fusion, a D-D reactor size would remain under an acceptable size. Indeed, a D-D power plant would be necessarily large and powerful, i.e. the net electric power would be equal to a minimum of 1.2 GWe and preferably above 10 GWe. A D-D reactor would be less complex than a D-T reactor as it is not necessary to obtain Tritium from the reactor itself. It is proposed the same type of reactor yet proposed by the author in a previous article, i.e. a Stellarator “racetrack” magnetic loop. The working of this reactor is continuous. It is reminded that the Deuterium is relatively abundant on the sea water, and so it constitutes an almost inexhaustible source of energy. Thanks to secondary fusions (D-T and D-He3) which both occur at an appreciable level above 100 keV, plasma can stabilize around such high equilibrium energy (i.e. between 100 and 150 keV). The mechanical gain (Q) of such reactor increases with the internal pipe radius, up to 4.5 m. A radius of 4.5 m permits a mechanical gain (Q) of about 17 which thanks to a modern thermo-dynamical conversion would lead to convert about 21% of the thermal power issued from the D-D reactor in a net electric power of 20 GWe. The goal of the article is to create a physical model of the D-D reactor so as to estimate this one without the need of a simulator and finally to estimate the dimensions, power and yield of such D-D reactor for different net electrical powers. The difficulties of the modeling of such reactor are listed in this article and would certainly be applicable to a future D-He3 reactor, if any.

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.

Proposal of a Deuterium-Deuterium Fusion/PWR Fission Hybrid Reactor

This article proposes to associate a Deuterium-Deuterium (D-D) fusion reactor with a PWR (fission Pressurized Water Reactor) in a hybrid reactor. Even if the mechanical gain (Q factor) of the D-D fusion reactor is below the unity and consequently consumes more energy than it supplies, due to the high energy amplification factor of the PWR fission reactor, the global yield is widely superior to 1. As the energy supplied by the fusion reactor is relatively low and as the neutrons supplied are mainly issued from D-D fusions (at 2.45 MeV), the problems of heat flux and neutrons damage connected with materials, as with D-T fusion reactors are reduced. Of course, there is no need to produce Tritium with this D-D fusion reactor. This type of reactor is able to incinerate any mixture of natural Uranium, natural Thorium and depleted Uranium (waste issued from enrichment plants), with natural Thorium being the best choice. No enriched fuel is needed. So, this type of reactor could constitute a source of energy for several thousands of years because it is about 90 more efficient than a standard fission reactor, such as a PWR or a Candu one, by extracting almost completely the energy from the fertile materials U238 and Th232. For the fission part, PWR technology is mature. For the fusion part, it is based on a reasonable hypothesis done on present Stellarators projects. The working of this reactor is continuous, 24 hours a day. In this paper, it will be targeted a reactor able to provide net electric power of about 1400 MWe, as a big fission power plant.

Experimental Instrumentation for Measurement of Reactivity Temperature and Voiding Effects at Zero Power Research Reactors

The paper describes the instrumentation for studying temperature and void reactivity effects that were developed at VR-I zero power reactor. Further are described its operational parameters, fields and ways of its utilization as well as issues connected to its implementation into the reactor core.

Development and Application of Maintenance Template in Pressurized Water Reactor Nuclear Power Plant

Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.

Case Study of Reactor Containment Building Construction in Nuclear Power Plant

It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. Through a case study, this study performs a pre-study for the reduction of construction duration in nuclear power plant project based on construction process of the RCB. The actual data of the case study have been collected and analyze the process and the external wall drawings of the RCB with construction practitioners. As a result of that, it is necessary to modularize the external wall form for equipment hatch and to extend the height of one layer of the external wall form to reduce the construction duration of RCB. The results of this study will be utilized to reduce construction duration of the nuclear power plant.

零功率反应堆研究进展与展望

零功率反应堆是开展反应堆物理研究重要的基础性实验平台,其结构简单,灵活可变,堆芯活性区能够针对特定的研究对象进行调整,以验证核设计软件与核数据库的准确性,获得堆芯物理重要参数实验数据,检验反应堆测量控制设备的有效性。本文简要描述零功率反应堆的特征与用途,回顾国内外研发历程,并介绍当前研究进展以及对未来的展望。

国际可移动小型堆开发与运输安全研究现状

可移动小型堆(TNPPS)作为小型模块化反应堆(SMR)的一种(一般指30 MW以下的SMR),当通过陆路、海运等方式运输后在运输工具上或者从运输工具卸载后能够运行和产生电能,可以满足偏远、局部区域的供热、用电需求,引起了各国的重视。本文简要介绍了国际可移动小型堆发展现状,重点对国际上不同类型可移动小型堆的特点、可移动小型堆发展面临的问题和挑战进行总结,并重点讨论了可移动小型堆运输安全相关问题及对策建议。

核电站堆腔混凝土辐照试验研究

作为核电站关键材料的堆腔混凝土,其安全服役是核电站长期稳定安全运行的前提条件之一。为了进行堆腔混凝土的中子辐照损伤机理研究,获得堆腔混凝土在中子辐照环境下的试验数据,本文建立了堆腔混凝土辐照试验方法,研制了辐照试验装置,并在研究堆中对其进行了加速辐照试验。结果表明:辐照试验装置设计合理,辐照试验指标满足试验要求,实现了两种规格多个混凝土试样的中子辐照。进一步的混凝土试样辐照性能研究结果表明:混凝土试样在平均快中子注量3.41×10^(18) cm^(−2)下辐照后,与辐照前相比,其外部形状未见明显差异,但试样颜色变化较大,并且出现一定的辐照肿胀和力学性能退化现象。

热管冷却反应堆系统研究进展和挑战

热管是一种高效的非能动热量传递元件,热管冷却反应堆核动力系统在多场景特种应用领域具备独特优势。本文概述了热管冷却反应堆特种核动力系统发展情况。首先介绍了热管冷却反应堆概念提出以及在海陆空天等领域的应用场景分析,同时总结了国内外典型堆型的发展现状。其次探讨了当前热管冷却反应堆面临的关键技术挑战,包括高性能材料研究、高性能热管研制、高效能量转换技术研究、设计分析技术研究。最后对未来发展趋势进行了分析和展望,强调了整体系统一体化研制、发电器件特性研究以及智能自主控制技术在热管冷却反应堆领域的重要性。本文的系统性总结将推动热管冷却反应堆技术的进一步发展,为未来特种核动力系统的应用提供重要支持。

池式钠冷快堆堆内自然循环余热排出设计研究

基于装量功率比约为1 t/MW的较典型池式钠冷大功率快堆的堆内自然循环设计,采用ERAC程序,对两组4种不同事故余热排出系统(DHRS)输入条件下,反应堆在紧急停堆后堆内的自然循环余热排出特性进行分析。结果表明,从DHRS启用到其对堆芯产生显著冷却效应,需要较长时间,在千秒量级。在该段时间内,堆芯余热的排出依靠反应堆固有的热工流体安全特性。随后,在堆内关键温度上升到限值之前启用DHRS带出池内热量,使堆内关键温度处于下降趋势即可满足安全要求;相比将独立热交换器(DHX)布置在冷池,将其布置在热池时,热池温度及主容器壁温相对要低,这有利于主容器的温度控制,其效果要优于布置于冷池。另外,不同布置会对堆芯盒内、盒间流流量产生影响,但总体上对堆芯的冷却效应影响不大;池式钠冷快堆余热排出设计中,要充分利用固有热工流体安全特性,降低对DHRS的时效性要求。可以考虑将全部的DHX都布置在热池,并缩小设备体积、降低散热功率设计值,或在不降低安全性的前提下选用其他更经济便捷的有效方式等,以此大幅降低余热排出设备投入成本,降低反应堆运行成本,提高经济性。本文研究结果可为我国后续的商用快堆、一体化快堆等池式液态金属堆的堆内自然循环设计提供重要参考。

10 kW低浓铀月表反应堆电源系统概念设计与分析

针对未来月球科研站所面临的能源需求问题,提出一种采用低浓化U-Mo合金燃料和热管冷却的月球表面反应堆电源系统的新型概念设计。低浓化U-Mo合金燃料技术成熟,在RERTR项目中被作为主要燃料而得到广泛使用。采用低浓化U-Mo合金燃料,能够有效降低核扩散的风险和简化相关部门的监管流程。反应堆的额定热功率为180 kW,输出电功率≥10 kW,设计寿命为10年。反应堆为Na热管冷却的快中子反应堆,采用控制鼓对堆芯反应性进行控制,堆芯热量通过Na热管传递到半Heusler型温差发电模块,废热由Hg热管辐射器翅片辐射到太空。基于蒙特卡洛程序RMC和CFD软件对反应堆堆芯物理及热工进行了初步计算和分析,针对温差发电器件的热电转换性能进行了优化设计,对热管失效等假想事故进行了分析评价。结果表明,该反应堆设计能够满足中子物理设计和临界安全的要求,在稳态工况下温差发电器件的热电转换效率能够达到7.60%,系统能够有效输出≥10 kW电功率,热管级联失效时能够满足热工安全要求。

双碳目标下高温气冷堆替代中小型火电的思考

碳达峰碳中和是我国一项重要的国家战略决策,“双碳”目标的提出给能源行业带来了深刻的变革,我国能源结构将进一步优化,电力行业必须实现低碳转型,新能源取代火电已成为必然。本文分析了新形势下火电企业面临的巨大挑战、发电规模受到限制、经营成本不断上升,尤其是中小型火电机组,面临着淘汰和关停的局面,而高温气冷堆具有固有安全性、发电效率高、用途广泛的显著特点,使其替代中小型火电成为可能。本文通过厂址适应性、技术可行性、经济可行性三方面,分析了高温气冷堆替代中小型火电的可行性,并从相关法律法规适用性、内陆厂址所带来的问题、开展公众沟通等方面提出建议。

超高功率电弧炉变压器和电抗器技术参数设计

电弧炉变压器主要作用是降低供电电压,产生大电流,在保证供电效率的情况下,适应电弧炉冶炼所需的电弧弧长,保证电弧炉冶炼过程的安全性。同时提供不同挡位的电压和电流,以满足冶炼过程需要。电抗器主要作用是增加线路的无功功率,降低短路电流和功率因数,稳定电弧燃烧。同时提供不同挡位的电抗,以满足冶炼过程不同工况的需要。变压器和电抗器技术参数的合理设计是保证电弧炉稳定冶炼的关键,依据电弧炉所需的冶炼效率及出钢量,系统分析电弧炉变压器和电抗器参数的设计原理。从分析可知,变压器的容量是基于电弧炉的通电时间及出钢量进行确定,而变压器的二次电压、额定电流及电抗器容量等参数是基于电弧冶炼稳定性的约束条件进行计算。计算结果与JB/T 9640-2014参考值进行对比分析,结果匹配度较好。

HPR1000： Advanced Pressurized Water Reactor with Active and Passive Safety

HPR1000 is an advanced nuclear power plant(NPP)with the significant feature of an active and passive safety design philosophy,developed by the China National Nuclear Corporation.On one hand,it is an evolutionary design based on proven technology of the existing pressurized water reactor NPP;on the other hand,it incorporates advanced design features including a 177-fuel-assembly core loaded with CF3 fuel assemblies,active and passive safety systems,comprehensive severe accident prevention and mitigation measures,enhanced protection against external events,and improved emergency response capability.Extensive verification experiments and tests have been performed for critical innovative improvements on passive systems,the reactor core,and the main equipment.The design of HPR1000fulfills the international utility requirements for advanced light water reactors and the latest nuclear safety requirements,and addresses the safety issues relevant to the Fukushima accident.Along with its outstanding safety and economy,HPR1000 provides an excellent and practicable solution for both domestic and international nuclear power markets.

大功率质子交换膜燃料电池电堆膜电极一致性研究

随着燃料电池应用领域的拓宽以及应用规模的不断扩大,大功率燃料电池电堆的需求也不断上升。大功率燃料电池电堆的电压一致性是衡量或影响电堆性能的重要指标。对某公司生产的65 kW大功率电堆进行了单电池一致性的研究,考察不同运行条件对于电堆一致性的影响,并对其产生的可能原因进行了深入的研究和讨论。研究结果表明:在额定功率和给定的运行条件下,电堆表现出了较好的一致性;电堆输出功率、电堆温度和温差、反应气体计量比、气体湿度等对电堆膜电极一致性均有较大的影响。其中输出功率、空气计量比、气体湿度对于电堆一致性的影响最为强烈,输出功率增高、空气计量比和空气湿度降低均会大幅度降低电堆的一致性。在研究结果的基础上,提出了保持大功率燃料电池电堆一致性的运行条件的建议。可为改善大功率电堆的设计和优化大功率电堆的运行条件,及促进我国燃料电池技术及产业的发展提供参考。

大型池式钠冷快堆热功率的计算和不确定度分析

反应堆的核功率无法直接测量,通过测量中子注量率的大小来表征核功率的大小,因此核电厂通常采用热功率刻度反应堆的核功率。池式钠冷快堆冷却剂系统采用钠-钠-水3个回路的布置形式,与压水堆存在较大差异,由于系统配置的不同,热功率的计算方法与压水堆也存在一定差异。该文分析了大型池式钠冷快堆热功率的计算和不确定度分析方法,用于反应堆核功率测量仪表的标定。计算结果表明,对于大型池式钠冷快堆,热功率计算的不确定度在事故分析初始功率的保守假设之内。堆芯热功率的计算结果主要取决于蒸汽发生器的功率计算,其余各项对最终计算结果的影响小于1%。

空间热管反应堆电源研究进展及展望

深空探测技术的发展对动力系统提出了更高的要求。传统的太阳能电源与化学电源的适用范围较小,环境适应能力不强,而微型核反应堆电源能量密度高,不依赖太阳光照,可应用于轨道运输、高轨探测多场景任务。在微型核反应堆电源技术路线中,热管冷却核反应堆电源因其系统设备极大简化、模块化设计,高可靠的全固态堆芯、非能动传热及瞬态响应迅速等特性,成为空间核反应堆电源最具可行性的路线之一。通过文献调研总结目前空间热管堆发展现状,从发展历史出发,梳理热管冷却核反应堆电源设计和理论研究,总结热管冷却核反应堆电源发展方向和关键技术。

并联电抗器补偿方案对500 kV海底电缆沿线过电压的分布影响

交流海底电缆是海上风电场交流输电技术的重要组成部分,但其充电电流问题限制了其在远距离、大容量海上风电中的应用。文中基于HYJQF41-F290/500 kV型号的单芯交流海底电缆,研究了该电缆在远距离、高压输电中不同工况下的沿线过电压分布情况。首先,建立该电缆的仿真模型,并对其参数进行修正。然后,针对并联电抗器的不同补偿度,对单端补偿、两端补偿等配置方案下并联电抗器的容量进行分析计算。最后,基于高压、远距离风电场模型中单相接地故障、合分闸等工况,对不同无功补偿方案下海底电缆沿线过电压的分布规律进行仿真研究。研究结果表明,海底电缆的最大过电压随着补偿方案的不同而出现在不同的位置。基于此,文中给出了在不同补偿方案下高压海底电缆选型以及变电站设备绝缘水平设计时过电压的选取方法,对工程实际具有一定指导意义。

42对棒环形布棒还原炉能量耗散研究

由于无法对大型还原炉环状硅棒排布方式对能耗的影响做出准确判断,对42对棒西门子多晶硅还原炉进行仿真模拟计算,建立两种环状硅棒布棒方式(三环排布、四环排布)的还原炉模型。通过对还原炉的炉行为及棒行为分析发现,硅棒直径加大能耗也会相应加大,且这种变化受炉内流场和温度场共同影响;同时,还原炉内内环硅棒辐射能耗随硅棒直径的增大呈先增大后减小的变化趋势,而排列在外环的硅棒辐射损耗则随硅棒直径的增大而不断增大。针对42对棒西门子还原炉内硅棒的排布方法,提出选用硅棒三环排布的方式可达到降低该还原炉总能耗的目的。