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实验包层模块(TBM)是聚变反应堆最重要的组件之一,作用是产氚和能量提取。锂陶瓷具有良好的化学稳定性、热机械性能、产氚性能以及可在更高温度下使用等特点,被认为是聚变堆包层最具吸引力的氚增殖剂材料。中国ITER-TBM设计方案采用了氦冷固态氚增殖剂(HCCB)TBM结构,其聚变环境下的辐照损伤行为可为中国HCCB TBM结构设计提供支持。针对固态氚增殖剂聚变中子辐照损伤问题,利用蒙特卡罗模拟,对比分析了Li_4SiO_4和Li_2TiO_3的中子辐照离位损伤和嬗变气体损伤。结果表明:在相同的服役时间下,Li_4SiO_4比Li_2TiO_3将产生更多的嬗变气体,且在高6 Li丰度情况下,其中子辐照损伤也更严重,会产生更高的损伤剂量和更大的损伤截面。但是,嬗变气体所造成的空位损伤Li_2TiO_3要比Li_4SiO_4严重;对两种陶瓷材料来讲,氦损伤效应均强于氚损伤效应。 相似文献
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FEB-E(Fusion Experimental Breeder)是聚变实验增殖堆的工程概要设计。FEB的主要目标是:(1)演示混合堆工程特性和裂变燃:抖和氚的增殖性能;(2)试验混合堆关键部件和聚变结构材料。环向场线圈TFC位于真空室及屏蔽层外侧,是FEB-E关键部件之一,其造价约占整个堆的40%。TFC由超导(Nb_3Sn)、绝缘体(聚酰亚胺)、稳定剂(Cu)和结构(316SS)等材料组成。由于TFC的超导、绝热和绝缘等材料易受来自堆芯聚变中子的辐照损伤,从而会严重影响混合堆的经济、稳定及安全运行,因此需要在等离子体堆芯与TFC之间设置一个屏蔽层把TFC所受的辐照损伤和核热沉积严格要求在允许范围以内。 相似文献
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吴宜灿 刘超 宋钢 王永峰 李桃生 汪建业 蒋洁琼 赵柱民 宋勇 胡丽琴 黄群英 李亚洲 王文 王志刚 王刚 季翔 王亮 王为田 于前锋 黄国强 程雄卫 王飞鹏 张思纬 李雅男 韩运成 宋婧 龙鹏程 FDS团队 《核科学与工程》2016,(1):77-83
强流氘氚聚变中子源HINEG(High Intensity D-T Fusion Neutron Generator)研发分两期:HINEG-Ⅰ为直流脉冲双模式,已成功产生中子强度1.1×10~(12)n/s的氘氚聚变中子,并实现连续稳定运行;HINEG-Ⅱ中子强度设计指标为10~(14)~10~(15)n/s量级,重点突破强流离子源和高载热氚靶技术。HNEG中子源可开展中子学方法程序与核数据、辐射屏蔽与防护、材料活化与辐照损伤机理和部件中子学性能等核能与核安全研究,同时也可在核医学与放射治疗、中子照相等领域拓展核技术应用研究。本文简要介绍HINEG总体设计方案与关键技术研究进展。 相似文献
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为测量聚变堆固态氚增殖剂堆内辐照氚增殖剂的产氚速率,除用常规电离室之外,本研究建立了Ne载气的高精度气相色谱在线检测分析方法,通过测量产氚回路中的氦产生量,验证系统中产生的氚量,从而为聚变堆固态包层产氚包层增殖剂材料辐照产氚性能提供一种新的产氚速率测量验证方法。本工作通过研制含有三个检测器、五个色谱柱的气相色谱分析系统,建立了Ne中微量4He、H2及杂质组分的色谱检测分析方法,并完成了实时在线检测的验证实验。结果表明,研发的色谱分析系统可实现高纯Ne中4He、H2及杂质组分的检测分析,H2、4He检测限可分别达到1.0×10-6、5.9×10-6,各组分含量及峰面积的相对标准偏差(sr)均小于5.0%(n=6),线性相关系数(r2)均大于0.99,说明检测方法重复性好。根据Ne中多组分气体的在线检测验证实验可知,单时段和多时段内的测量重复性均较好,可为辐照产氚考核系统中的产氚速率验证提供分析手段,进而为正式入堆得到辐照数据和氚衡算提供技术支持。 相似文献
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托卡马克聚变堆的主要发展方式包括混合堆、纯聚变堆。关于托卡马克聚变堆氚自持的研究,国内外主要采用平均滞留时间方法进行研究,并且针对聚变功率较低的混合堆的氚自持研究较少。本工作采用更符合实际的积分分析方法分析了混合堆、纯聚变堆氚自持的启动氚量、氚增殖比(TBR)要求。研究结果表明:启动氚量、备用氚量与聚变功率具有线性关系,所需TBR与聚变功率呈反比例关系;混合堆聚变功率较低,所需TBR较高,工程实现所需TBR挑战较大,需要通过限制长期氚滞留量以降低所需TBR要求;纯聚变堆聚变功率高,所需TBR较低,工程实现所需TBR挑战较小,但备用氚需求达数十千克,应考虑氚系统的冗余设计或提高氚系统的可靠性、可维护性,以降低备用氚的使用规模;运行因子是聚变堆的一个重要设计指标,在此着重分析了运行因子对所需TBR的影响,并重新定义了一个聚变堆氚自持的关系式,以突出运行因子对氚自持的重要影响。 相似文献
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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。 相似文献
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采用通用蒙特卡罗粒子输运程序MCNP/38计算低环径比Tokamak(紧凑环或球形环)聚变堆第一壁及中心导体上的中子壁负荷分布和核热沉积分布,并与常规Tokamak堆第一壁上中子壁负荷分布和核热沉积分布进行比较、分析。结果表明,在中子壁负荷归一化为1MW/m2时,与常规Tokamak相比,在低环径比Tokamak堆第一壁及中心柱表面上中子壁负荷分布峰值并不比常规Tokamak堆第一壁上的峰值高,而且低于低环径比Tokamak堆整个第一壁上的平均值,而中心柱上的核热沉积峰值稍高于常规Tokamak堆第一壁上的核热沉积峰值,但对较高中子壁负荷情况,中心导体柱上的核热沉积和辐照损伤仍可能是比较严重和值得特别研究的问题。 相似文献
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J. Jordanova U. Fischer Y. Poitevin N. Nikolova-Todorova 《Fusion Engineering and Design》2006,81(19):2213-2220
Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a 6Li enrichment of 90%. Assuming a 6Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard. 相似文献
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《Fusion Engineering and Design》2014,89(4):412-425
A high-β spheromak reactor concept has been formulated with an estimated overnight capital cost that is competitive with conventional power sources. This reactor concept utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt (FLiBe) blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER-developed cryogenic pumping systems were implemented in this concept from the basis of technological feasibility. A tritium breeding ratio (TBR) of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. 相似文献
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次临界能源堆物理性能初步分析 总被引:2,自引:1,他引:1
次临界能源堆(SER)是由托卡马克聚变源驱动的聚变裂变混合堆。SER以天然铀为燃料、水为冷却剂,主要目标是生产电能。本工作建立了次临界能源堆环形圆柱模型,利用蒙特卡罗输运和燃耗计算程序,比较了燃料区不同构型对keff、M、TBR和燃料增殖比等参数的影响,针对均匀模型进行中子源效率与聚变源强、功率分布与能谱、初步燃耗、寿期末停堆衰变热和卸载燃料放射性等物理性能分析。计算结果表明,该模型能满足能量倍增大于6、氚自持、较长时间不换料等设计目标。研究结果为下一步开展SER安全分析提供了基础。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(4):301-302
The poloidal distribution of the first wall 14 MeV neutron flux and the tritium breeding ratio in a Tokamak fusion reactor were calculated using Monte Carlo method. The poloidal distribution of the 14 MeV neutron flux in the first wall was found to be quite different from that of the primary incident flux. The tritium breeding ratio calculated by the Monte Carlo method became about 5% larger than the value obtained from SN transport calculations. 相似文献
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Robert W. Conn Gerald L. Kulcinski Charles W. Maynard 《Nuclear Engineering and Design》1976,39(1):5-44
UWMAK-II is a conceptual design study of a low ß, circular Tokamak fusion power reactor. The aim of the study has been to perform a self-consistent analysis of a probable future fusion power system based on the philosophy that design decisions, wherever possible, should be conservative and should be based on present technology. As such, this system will not be the smallest, the least expensive, or the optimum Tokamak reactor. Rather, it represents a feasible system which we use to assess the technological problems uncovered and to examine possible solutions. The plasma is designed to generate 5000 MW(th) during a pulse and 1709 MW(e) continuously based upon a burn cycle with a 90 min burn and a 6.5 min rejuvenation period. The plasma carries a current of 14.9 MA and is designed with a double null poloidal divertor for impurity control and particle pumping. In addition, a low Z liner in the form of a carbon curtain is included to eliminate any source of high Z impurities. Plasma heating to ignition involves the use of neutral beam heating for a 10 sec period during which 200 MW of 500 keV deuterium atoms are injected into the plasma.The blanket design employs helium cooling and the solid lithium-bearing compound, lithium aluminate (Li2Al2O4) for breeding tritium. The structural material is 316 stainless steel and beryllium is used as a neutron multiplier. The neutron radiation environment produces radiation damage that considerably influences blanket and system performance. The most significant effect is the loss of ductility which appears to limit the usable lifetime of the blanket first wall to about 2 yr at a 14 MeV neutron wall loading of 1.16 MW/m2. The solid breeder blanket minimizes the tritium inventory but because of the low fractional burnup in the plasma and the need for roughly a one day reserve of fuel, the inventory is 17.7 kg. Induced radioactivity levels in the structure are of the order of 1 Ci/W(th) at shutdown after two years of operation. The main contributors to the activity are
) and
). Afterheat levels are slightly above 1% of thermal power but the afterheat power density is low, less than 0.1 w/g. The power cycle involves a He---Na intermediate heat exchanger followed by a sodium—steam system. The sodium intermediary is used to minimize tritium leakage through the power cycle and to provide a working fluid for thermal energy storage such that continuous electrical output is produced despite a pulse plasma cycle. A materials resource study has been completed for a UWMAK-II type system and beryllium appears to present a particular problem with regard to adequate resources. Other materials that could present problems of procurement include chromium and nickel. A preliminary economic analysis has been carried out to identify major cost areas and this is described. 相似文献
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托卡马克(Tokamak)聚变装置中子学分析中,聚变中子源描述是重要的输入参数,其准确性直接影响分析结果的可靠性。通过分析ITER和欧洲聚变示范堆(EU DEMO)中子学分析中所采用的聚变中子源模型,提出了一种完整描述Tokamak中L-mode、H-mode等离子体的D-D、D-T聚变中子源的数值模型。在中国聚变工程实验堆(CFETR)的工程集成设计平台上,编写了基于蒙特卡罗算法的程序SCG求解该数值模型,实现了读取(零维)等离子体参数、输出可供典型中子学软件MCNP直接读取的中子源定义文件的功能。以CFETR氦冷球床包层的中子学分析模型为基准,在相同的L-mode等离子体D-T聚变工况下,相较于采用EU DEMO源子程序,采用本模型计算得到的中子壁负载差异最大值为2.02%,包层氚增殖率差异为0.18%,全堆能量增益因子的差异为0.23%。结果表明,本模型与其他源描述的差异较小,可应用于CFETR的中子学分析。 相似文献
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With the tremendous surge in the usage of radioactive materials in industry, education and research, medicine and other fields, it becomes a concern to protect the working personnel and common people around, from hazardous radiation leakages that may seriously affect their health. Among the different types of radiation, gamma and neutron radiations require adequate shielding. There have been several attempts to develop newer concretes and evaluate their neutron radiation shielding characteristics. In the present study, an attempt has been made to study the effect of varying the mix parameters and hence the resulting total hydrogen content on the neutron radiation shielding characteristics of Latex Modified Concrete (LMC) mixes. The experiments are planned in such a way that the hydrogen content of the mixes is varied by controlling the mix parameters i.e., cement content, water/cement ratio and polymer/cement ratio of LMC mixes. The results are statistically analyzed. It is found that definite improvements could be achieved in neutron radiation shielding characteristics of LMC mixes as compared to ordinary concrete, with the increase in hydrogen concentration effected by changes in mix parameters. 相似文献