CFETR第一壁及赤道面外包层中子辐照损伤初步分析

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。     

~(241)Am-Be源在煤中所形成中子场的参数拟合

研究了7种煤中主要元素对241Am-Be中子源在煤中形成中子场的影响,给出了描述中子场中快中子和热中子数量变化曲线的经验公式和拟合参数。     

冷中子源封闭的两相热虹吸系统自稳特性研究

提出包含缓冲罐在内的封闭两相热虹吸自然循环回路的自稳模型.在缓冲罐与冷凝器间连接管的绝热和瞬间温度平衡两个条件下,研究其自稳特性,并与日本京都大学提出的模型进行了对比,提出提高自稳特性的措施.     

基于GDT的14MeV中子源初步设计研究

为满足聚变材料测试对D-T聚变中子源的需要,本文首先根据国际上对用于聚变材料测试的中子源的要求给出设计目标,然后基于Gas Dynamic Trap(GDT)装置的实验进展,提出了基于GDT装置的14MeV中子源的设计初步方案,并建立了GDT中子源的物理模型。计算分析给出了两套中子源参数初步方案,其中FDS-GDT2中子壁负载为2MW/m2,可用于聚变材料的测试。     

利用LR 115 2B研究7.4GeV质子束轰击铅靶的散裂中子源外表面中子分布

利用LR 115 2B探测器研究了8cm× 2 0cm的铅靶在 7.4GeV质子的轰击下 ,散裂中子在实验装置外表面的分布 ,对实验结果进行了模拟解析 ,同时利用中子活化法验证了LR 115 2B探测结果及理论模拟结果利用LR 115 2B探测器研究了8cm× 2 0cm的铅靶在 7.4GeV质子的轰击下 ,散裂中子在实验装置外表面的分布 ,对实验结果进行了模拟解析 ,同时利用中子活化法验证了LR 115 2B探测结果及理论模拟结果     

多时空通道中子注量率密度数据采集系统方案设计

介绍了用在加速器驱动核能系统(ADS)次临界原理验证装置上的微机化多时空通道中子注量率密度数据采集系统设计方案．该数据采集系统用于采集多个空间位置的中子注量率密度水平随时间变化的数据。     

The effect of wall loading limitations and choice of beta on the feasibility of advanced fuel fusion reactors

This paper investigates the effect of wall loading limitations and choice of plasma stability index beta on the feasibility of advanced fuel fusion reactors. Two new conceptual tools are introduced to facilitate this analysis: the “effective reactivity,” which includes all of the reaction-relevant parameters that determine the fusion power density, and the “critical radius,” which is the maximum allowable minor radius of a fusion reactor, beyond which the power generated in the plasma will exceed allowable loadings of radiant energy or neutrons on the first wall. It is shown that if high beta (greater than 0.2) fusion reactors are feasible, the high reactivity of the DT reaction cannot be fully exploited because of wall loading limitations. In addition, some high beta reactors with advanced fuels are also found to be wall loading limited, and to have excess reactivity, which can be traded off for lower magnetic fields, longer particle containment times, etc. Under certain circumstances, the reduced materials problems associated with some advanced fuels may outweigh the reactivity advantage of the DT reaction, and make one of them the reaction of choice for high beta fusion reactors.     

Effect of mix parameters and hydrogen loading on neutron radiation shielding characteristics of latex modified concrete mixes

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.     

Peaked density profiles and MHD activity on FTU in lithium dominated discharges

Density profiles become broader, as the line averaged density is increased. At higher density, change of the trend is associated with the appearance of the MARFE. Lithium coated wall extends the maximum density accessible at low current (well above the Greenwald density limit) and can produce profiles with very high density gradient. At higher current the effect on the density limit can be exceeded only if the magnetic field is raised too. A comparison of successive similar discharges before and after wall conditioning with lithium showed a reduction of the MHD activity.     

靶件内中子注量率分布计算方法研究

选择ANISN作为实验靶件内中子注量率分布计算的程序，编制辅助程序输入混合材料截面。计算得到延时水箱附近的中子注量率，与测量数据作对比。计算得到靶片自屏因子，并与2000年实验数据对比。确认计算方法可行后，计算得到实验靶件内热中子注量率分布数据。     

Count-loss effect in subcriticality measurement by pulsed neutron source method, (I) investigation on count-loss effect in determination of neutron decay constant

The count-loss effect in determination of neutron decay constant by pulsed neutron source method was investigated. It was found that overestimation of neutron decay constant due to count-loss effect is seen while underestimation appears superiorly as the intensity of pulsed neutron source is getting higher. It was further demonstrated that the well-known count-loss correction procedures are not effective for overestimation although they suppress underestimation. Therefore, the pulsed neutron source method should be modified so as to have robustness against the count-loss effect.     

Correction of count-loss effect in neutron correlation methods that employ single neutron counting system for subcriticality measurement

It is well known that the count-loss effect produces serious problems for the neutron correlation methods that employ a single neutron counting system, e.g. the variance-to-mean and the auto-correlation methods, since it deteriorates the information extracted from the subcritical reactor system. New formulae of the variance-to-mean and auto-correlation methods were hence obtained on the basis of a rigorous theoretical approach for treating the count-loss process. It is expected that the present formulae work better than conventional ones for determination of the neutron decay constant.     

用散射中子测量管道油垢厚度的实验与模拟

由~(241)Am-Be中子源、锂玻璃探测器和微机多道谱仪等组成的实验装置,测量了不同石蜡(模拟油垢)厚度的散射中子计数.同时用蒙特卡罗模拟,分别对~(241)Am-Be中子源、~(252)Cf中子源和14MeV中子源,计算得到了相似几何条件下,对应于不同石蜡(模拟油垢)厚度的散射中子计数.在模拟计算中,考虑了锂玻璃探测器的探测效率.对实验和模拟计算的结果进行了比较和讨论.     

核临界安全中的源倍增法研究

文章对核临界安全研究中通常采用的现场测量技术———源倍增法进行研究。从有源扩散理论出发,导出了与keff不同的有源次临界中子有效增殖因子ks的表达式,并在次临界系统上进行了验证研究。验证实验研究证实了所导出的ks 的正确性。源倍增法测量的参数实际上是次临界系统在外源作用下的有源次临界中子有效增殖因子ks,而不是以往的中子有效增殖因子keff,这就解决了长期困扰人们的有关源倍增法测量的参数问题。文章讨论了ks 与keff间的差别和关系以及它们对核临界安全的影响。     

Experimental study of neutron counting in a zero-power reactor driven by a neutron source inherent in highly enriched uranium fuels

Even a zero-power reactor core containing highly enriched uranium has a weak neutron source inherent in uranium 235, and consequently, a neutron counter placed closely to the core without external neutron source registers a certain counting rate. The study of the counting is very important for zero-power reactor physics experiments with a high precision. In this experimental study, first, at a shutdown state of the UTR-Kinki reactor without start-up neutron source, a pulse height distribution of output signals from a neutron proportional counter was measured to confirm that these signals resulted from neutron detections. At several subcritical states of the UTR, then, the Feynman-α analysis was carried out to confirm that the neutrons detected by the counter must be fission neutrons multiplied by fission chain reactions. The correlation amplitude measured in the Feynman-α analysis was much higher than that measured in a previous drive by start-up source. Further, it was also confirmed that the subcriticality dependence of neutron counting rate followed the source multiplication formula. This feature indicated that the one-point model was very successful in the subcritical range including the shutdown state.     

Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k_s, external source efficiency φ^*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton–tungsten source in hard and soft neutron spectra cores and 14 MeV D–T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6–13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.     

Geant4 simulation of 238U(n,f) reaction induced by D-T neutron source

Knowledge of actinides(n,f) fission process induced by neutron is of importance in the field of nuclear power and nuclear engineering,especially for reactor applications.In this work,fission characteristics of~(238)U(n,f) reaction induced by D-T neutron source were simulated with Geant4 code from multiple perspectives,including the fission production yields,total nubar,kinetic energy distribution,fission neutron spectrum and cumulative γ-ray spectrum of the fission products.The simulation results agree well with the experimental nuclear reaction data(EXFOR) and evaluated nuclear data(ENDF).Mainly,this work was to examine the rationality of the parametric nuclear fission model in Geant4 and to direct our future experimental measurements for the cumulative fission yields of ~(238)U(n,f) reaction.     

Application of different delayed neutron data sets to the analysis of rod drop experiments on VVER cores

Rod-drop experiments performed on VVER-440 reactors are analysed using different delayed neutron data sets. It is found that for these experiments the 8 group data sets of Spriggs et al [6] give results in agreement with the original 6 group sets of Keepin on which they are based. The results obtained using Brady and England data [3] are markedly different. Sensitivity studies are made to analyse these differences.