Design methodology for insulating and cooling of a small modular reactor head by high temperature structural-thermal-fluid analysis

The design of a sodium-cooled fast reactor (SFR) head can be complicated due to its shape and functions. The head is a component placed in the pressure boundary to shield nuclear radioactive radiation. At the same time, it needs to seal the reactor vessel, support penetrating components, and minimize heat losses. This paper presents a new insulating and cooling design concept of a small SFR head. For a new design, this study shows a comprehensive design approach considering fluid-thermal-structural computations. The interactive design approach refers to dependent simulation steps of three-dimensional (3D) thermal-structural, one-dimensional (1D) heat-transfer, and 3D computational fluid dynamics (CFD) analysis. This multi-domain approach was applied to the head of the large sodium integral effect test facility called sodium test loop for safety simulation and assessment (STELLA-2). And the STELLA-2 head design was proposed as a thick plate with a sandwich type of insulation, cooling the perimeter annulus of the round head-top surface. For the structural design, the ASME design code was utilized, and the head temperature of 346?°C was calculated as its initial design temperature target. In an axial heat-transfer mode from the in-vessel to the head, a 1D finite element model gave 57 and 75 mm insulation thicknesses with a thermal conductivity of 0.07 W/m/K. The cooling effectiveness of the proposed head design was shown through a commercial CFD package.     

Three pass core design proposal for a high performance light water reactor

The paper describes a novel core concept for a nuclear reactor cooled with supercritical water, in which the coolant is heated up from 280 °C at the reactor inlet to 500 °C at the outlet in four steps: a first heat-up step is provided by heat transfer from fuel assemblies to the moderator water in gaps and moderator boxes, a second step is foreseen in a central “evaporator” and two further steps in a first and a second superheater surrounding it. The coolant flow scheme includes upward and downward flow through the core with intermediate mixing in chambers above and below the core to eliminate hot streaks. A preliminary single channel analysis, concentrating on an average flow channel and on the hottest one only, indicates that such core design can match the limits of cladding materials available today. Even though the resultant pressure drop of the coolant will be higher than usual, it is expected that the assembly boxes can be designed with acceptable deformations.     

聚变高温制氢反应堆概念设计研究

在深入分析相关领域研究发展状况的基础上,提出了具有较好技术可行性的聚变高温制氢反应堆概念(称之为FDS-Ⅲ),包括具有先进等离子体物理和技术水平的聚变堆芯、先进高温锂铅包层(HTL)、可减少热流分布密度的"垂直靶板"偏滤器以及相应的功率转换系统。尤其是提出了HTL包层新概念,其特点是选用技术基础相对成熟的低活化铁素体/马氏体钢作结构材料,在锂铅流道中使用可耐高温的多层流道插件,实现约1000℃的出口温度,可应用于制氢。初步性能分析表明FDS-Ⅲ制氢堆及其包层概念具有较好的技术可行性。     

聚变发电反应堆概念设计研究

在广泛分析聚变能相关领域研究发展状况和国际热核聚变实验堆(ITER)物理与技术基础上,提出了一个考虑了技术可行性的聚变发电反应堆概念(称之为FDS Ⅱ)。这个概念具有ITER参数适量外推的等离子体物理与技术水平的聚变堆芯和具有发展潜力的液态锂铅氚增殖包层,在对这个概念进行中子学、热工水力学、力学、安全与环境影响和经济学等一系列计算分析的基础上,给出了初步的概念设计和进一步设计优化的共性原则。     

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...     

聚变发电反应堆双冷锂铅包层热工水力学设计与分析

依据结构设计和中子学计算结果给出了聚变发电反应堆FDS-Ⅱ双冷锂铅(DLL)包层热工水力学设计方案。采用数值计算软件对液态金属增殖区LiPb流场和第一壁热-结构等进行了模拟,并对功率转换系统的热效率进行了计算。通过分析评估,证实该包层热工水力学方案能较好地实现FDS-Ⅱ聚变发电反应堆预期目标。     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

Lithium nitrate as a fusion reactor coolant fluid?: A thermochemical assessment

In connection with possible application as a coolant fluid in fusion reactors, molten LiNO₃ and LiNO₃/LiNO₂ mixtures, and their mixtures with Na/K nitrates, have been evaluated with respect to their high temperature stability, their ability to reversibly dissolve and release tritium and, in a very limited sense, their corrosiveness toward structural alloys. The results of the primarily thermochemical evaluation indicate that with respect to thermal/radiolytic decomposition and corrosivity LiNO₃/LiNO₂ may be suitable for application up to R 700^°K; however, with respect to reversibility/irreversibility of tritium release, it appears unsuitable at all likely operating temperatures (R 675–800°K).     

Design of the material performance test apparatus for high temperature gas-cooled reactor

Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor （HTGR）. To solve the problem, a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the HTGR. The apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the apparatus. Specially designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field. A typical experimental process was analyzed in the paper, which lasted 76 hours including seven stages. Experimental results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and verified. Test temperature in the apparatus could be elevated up to 1600℃, which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test reauirements of materials used in the reactor.     

聚变发电反应堆双冷液态锂铅包层氚增殖中子学分析研究

针对聚变发电反应堆(FDS Ⅱ)双冷液态锂铅(DLL)包层进行了中子学设计与分析,设计主要的原则是满足聚变堆的氚自持,并在此基础上,分析计算DLL包层核热分布。中子学一维优化分析使用的程序是自主开发的多功能中子输运/燃耗/优化程序VisualBUS1.0以及相应的数据库HENDL1.0/MG。基于二维模型进行校核计算所使用的程序为MCNP4C,相应的数据库为FENDL 2/MC。     

Assessment of radioactive wastes from a DCLL fusion reactor: Disposal in El Cabril facility

Under the Spanish Breeding Blanket Technology Programme TECNO_FUS a conceptual design of a DCLL (Dual-Coolant Lithium–Lead) blanket-based reactor is being revised. The dually cooled breeding zone is composed of He/LiPb and SiC as material of the liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer) for the blanket and austenitic steel (SS316LN) for the vacuum vessel (VV) and the cryostat.In this work, radioactive wastes are assessed in order to determine if they can be disposed as low and intermediate level radioactive waste (LILW) in the Spanish near surface disposal facility of El Cabril. Also, unconditional clearance and recycling waste management options are studied.The neutron transport calculations have been performed with MCNPX code, while the ACAB code is used for calculations of the inventory of activation products and for activation analysis, in terms of waste management ratings for the options considered.Results show that the total amount of the cryostat can be disposed in El Cabril joined to the outer layer of both VV and channel inserts, whereas only concrete-made biological shield can be managed through clearance and none of the steels can be recycled. Those results are compared with those corresponding to French regulation, showing similar conclusions.     

Requirements for high temperature reactor fuel particle design assessment

In the framework of a large Research and Development programme devoted to High Temperature Reactors (HTR) and set up in the CEA from 2000 on, we will address ourselves to the issue of coated fuel performance and design. Although HTR fuel main features have been established for a long time, we need today to reassess the fuel design to make sure that it meets the requirements linked to the most recent projects of High Temperature Reactors. Thus, in collaboration with Framatome and in connection with the Gas Turbine - Modular Helium Reactor (GT-MHR) international project, we are planning to perform parametric thermal and mechanical studies, regarding different particle design options (kernel diameter, layers composition and thickness) and seeking optima concerning particle leak tightness and fission product retention. But to initiate such studies, we have first of all to define the design bases and the requirements for HTR fuel, in terms of kernel composition (fissile element, oxide stoechiometry, enrichment), particle and compact geometry (dimensions, particle volume fraction in the graphite matrix), power density, cooling gas temperature and irradiation conditions (burnup, fast fluence).     

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.     

中国低活化马氏体钢CLAM在液态锂铅中腐蚀的初步实验研究

液态金属锂铅包层是最具发展潜力的聚变堆包层之一,其首选结构材料为低活化铁素体/马氏体钢,而它与液态锂铅的相容性是聚变堆材料研究领域的关键问题之一.本文介绍中国低活化马氏体钢CLAM在液态金属锂铅回路DRAGON-1热对流工况下的实验情况及500 h 480 ℃下初步腐蚀实验结果,并与同样工况下316L奥氏体钢腐蚀结果进行了对比分析.结果显示CLAM钢与液态锂铅的相容性优于316L钢.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

Control rod drive for high temperature gas cooled reactor

This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.     

聚变反应堆含氚废水处理系统的设计

以联合电解催化交换-气相色谱(CECE-GC)为技术路线基础,对聚变反应堆(International thermonuclearexperimental reactor,ITER)含氚废水处理系统(Water detritiation system,WDS)进行了总体设计和主要子系统的设计.与目前的重水提氚演示系统相比,ITER-WDS的不同之处在于不使用氢氧复合器,不采用碱性电解池而使用固体聚合物电解池(Solid polymer electrode,SPE),增加了Pd/Ag膜渗透系统进行氚的回收.     

Design and licensing strategies for the fluoride-salt-cooled,high-temperature reactor (FHR) technology

Fluoride-salt-cooled, high-temperature reactor (FHR) technology combines the robust coated-particle fuel of high-temperature, gas-cooled reactors with the single phase, high volumetric heat capacity coolant of molten salt reactors and the low-pressure pool-type reactor configuration of sodium fast reactors. FHRs have the capacity to deliver heat at high average temperature, and thus to achieve higher thermal efficiency than light water reactors. Licensing of the passive safety systems used in FHRs can use the same framework applied successfully to passive advanced light water reactors, and earlier work by the NGNP and PBMR projects provide an appropriate framework to guide the design of safety-relevant FHR systems. This paper provides a historical review of the development of FHR technology, describes ongoing development efforts, and presents design and licensing strategies for FHRs. A companion review article describes the phenomenology, methods and experimental program in support of FHR.     

Design study of molten-salt-type reactor for powering space probes and its automated start-up

Space probes exploring the deep space need a noticeable amount of electricity for powering their instrumentation (5–8 kWe). The solar battery, however, is not able to afford this because of the extended distance between the probe and the sun. Furthermore, the radioisotope thermoelectric generator (RTG) may be extraordinarily sizable for this purpose. On the contrary, installation of nuclear reactor on the probes seems to be promising for the reasons such as its small mass, stability, and its high power density. Especially, the deep-space sample return mission has an essential demand for a nuclear reactor typically with the following specifications: 10-year reactor employment time, total system mass less than 500 kg, and electricity output more than 8 kWe. High power density is the key to realize these requirements. The thermal electricity conversion efficiency of a space reactor increases in accordance with the rise of the core temperature. Thus, in the present study, we propose a use of the molten-salt fuel considering that this type of fuel can achieve high core temperature over 1000 K. In addition to the space reactor for the space exploration, autonomic reactor control is extremely desirable because real-time control from the earth is difficult because of the long time lag of mutual communication. The molten-salt space reactor proposed in the present paper satisfies all the above essential demands primarily owing to the high-temperature operation availability. In addition, we concluded that the automatic reactor start-up in its orbit is feasible by virtue of introduction of novel reactivity-control devices proposed by Kambe (Kambe, Sato, Tsunoda. Space Nuclear Conference; 2007 June 24–28; Boston, MA. p. 36–45).