Recent progress on the Tokamak Fusion Test Reactor

The deuterium-tritium (D-T) experiments on the Tokamak Fusion Test Reactor (TFTR) have yielded unique information on the confinement, heating and alpha particle physics of reactor scale D-T plasmas as well as the first experience with tritium handling and D-T neutron activation in an experimental environment. The D-T plasmas produced and studied in TFTR have peak fusion power of 10.7 MW with central fusion power densities of 2.8 MWm^–3 which is similar to the 1.7 MWm^–3 fusion power densities projected for 1,500 MW operation of the International Thermonuclear Experimental Reactor (ITER). Detailed alpha particle measurements have confirmed alpha confinement and heating of the D-T plasma by alpha particles as expected. Reversed shear, highl _i and internal barrier advanced tokamak operating modes have been produced in TFTR which have the potential to double the fusion power to 20 MW which would also allow the study of alpha particle effects under conditions very similar to those projected for ITER. TFTR is also investigating two new innovations, alpha channeling and controlled transport barriers, which have the potential to significantly improve the standard advanced tokamak.     

D-T中子源束斑尺寸测量技术及应用

D-T中子发生器作为高能中子源,已用于特种核材料的有源探测技术中,利用时间符合法可实现被测物体的多模式成像,而中子源束斑尺寸是影响成像位置分辨的一个重要的因素。因此,结合D-T反应的特点和实际应用环境,开发了n-α关联符合测量确定中子源束斑尺寸的方法。利用该方法对小型移动中子发生器的束斑尺寸进行了测量,获得的束斑尺寸为(2.8±0.9) mm,与用CCD相机直接观测得到的约3 mm的测量结果一致,证明了该方法测量束斑尺寸的可行性。该方法也可用于辅助D-T中子源调束和关联粒子成像实验过程中的束斑监测。     

An FEL-based microwave system for fusion

This paper describes designs for 280-GHz and 560-GHz microwave sources based on free electron lasers (FELs). These 10-MW units are based on technology developed over the last 5 years. A first demonstration of high-average-power microwave production with an FEL system is expected in the Microwave Tokamak Experiment (MTX) facility. This paper gives details on the design and construction of that 250-GHz, 2-MW system and discusses specific applications for the Compact Ignition Tokamak (CIT).     

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.     

中国聚变工程试验堆包层的核热耦合效应研究

本文以中国聚变工程试验堆（CFETR）的氦冷固态包层和水冷固态包层为研究对象,基于蒙特卡罗程序MCNP和计算流体力学程序FLUENT,利用3D-1D-2D耦合方法和伪材料方法,分别对200 MW的氦冷固态包层和水冷固态包层及1.5 GW的水冷固态包层方案进行了核热耦合计算分析。研究结果表明,金属铍的热散射效应和轻水密度是聚变包层核热耦合效应的主要来源,核热耦合效应对氦冷固态包层的影响可忽略,对水冷固态包层的氚增殖比和温度分布有一定程度的影响。     

HFETR运行寿期和燃料元件装载量优化研究

文章建立了中子转换比与运行寿期之间关系的分析模型,设计出不同运行寿期的实际堆芯并进行计算,研究了⁶⁰Co产量和中子转换比随高通量工程试验堆（HFETR）运行寿期的变化规律。同时,通过对新燃料元件堆芯的研究找出燃料元件装载量对⁶⁰Co产量和中子转换比的影响,采用点堆模型分析平衡堆芯下HFETR的燃料元件装载量。该优化研究的目的在于为HFETR堆芯装载和运行方式优化提供参考以提高其运行的经济性。结果表明,HFETR运行寿期设计为25 d较佳,在此寿期下的平衡堆芯燃料元件理想装载量为70盒。     

基于可移动紧凑加速器D-T中子源的热中子层析实验研究

中子照相是十分重要的无损检测方法之一,尤其是针对含氢材料、同位素等的无损检测,中子照相技术具有其他射线成像不可比拟的优势。中国工程物理研究院核物理与化学研究所基于紧凑型D-T中子源,研发了可移动中子成像检测仪,成功实现了热中子照相和快中子照相实验检测。为确定基于该装置开展热中子层析检测的可行性,本文进行了数值模拟计算,利用该仪器开展了针对轻重材料模拟件的热中子层析成像实验,利用采集的181幅投影图像,在图像信噪较低和采集幅数较少条件下,成功重建了铝和聚乙烯材料包裹下的0.2 mm直径的钆丝。     

HFETR辐照靶件设计程序开发

针对高通量工程试验堆（ＨＦＥＴＲ）燃料材料考验的辐照靶件设计，开发了ＧＥＮＧＴＣ－Ｃ程序，可用它进行辐照靶件设计和温度分布计算，程序可对多层包壳材料传热，间隙层（液体或气体介质）传热，气体介质层传热进行了计算，同时，对热导膜率随温度的变化和靶件结构材料的几何尺寸因热膨胀的改变进行修正，原程序由美国ＯＲＮＬ编制，见长于结构严谨流畅和良好的适用性，因此，ＧＥＮＧＴＣ－Ｃ是在原程序的基础上并结合ＨＦＥＴ     

HFETR运行十年辐射工作人员个人内照射监测与评价

本文给出了高通量工程试验堆（ＨＦＥＴＲ）运行十年的辐射工作人员内照射监测结果，并按照ＩＣＲＰ３０号出版物和ＧＢ８７０３－８８推荐的模式和参数，估算出个人摄入量和待积有效剂量当量，并对十年监测结果进行了分析和评价。十年最大摄入量和最大待积有效剂量当量分别为３．７×１０＾５Ｂｑ和１０．３７ｍＳｖ，证明ＨＦＥＴＲ的运行对工作人员个人内照射的影响是很小的。     

Effects of Particle Confinement and Recycling on Thermally Stable Regions in D-T Tokamak Plasma

The water accumulation phenomena in the upper plenum of a PWR was investigated by using the data of the Cylindrical Core Test Facility (CCTF) test.

When the liquid inventory in the upper plenum is small, the liquid carryover rate from the upper plenum was governed by the liquid inlet flow rate. After some amount of the water was accumulated in the upper plenum, the liquid carryover rate was governed by the steam up-flow rate and the liquid inventory in the upper plenum. The liquid carryover rate was higher with the higher steam up-flow rate and the larger liquid inventory. This trend on the liquid carryover rate was also observed in small scale model test.

Based on the CCTF data, the empirical correlation on the liquid carryover rate was obtained.     

岷江试验堆核探测器布置的改进

岷江堆(MJTR)仪表控制系统用的核探测器．因距离堆芯较近．受较强的中子照射和^y辐射而使其使用寿命缩短;另外．随着堆内辐照孔道的孔径扩张、中心位置的变更及辐照样品体积的增大．使探测器处的中子信号受样品进出堆的干扰变大．对反应堆的安全运行造成不利影响。为了解决上述问题,对MJTR核探测器的布置进行了改进：本文介绍了在改进设计中．采取的措施以及设计的原则、方法和结果。改造完成后的各项调试结果表明．重新设计的探测器孔道布置和功能分配合理．完全满足设计要求。     

基于RELAP5的CFETR氦冷陶瓷增殖包层模块热工安全分析

本文对中国聚变工程实验堆(CFETR)氦冷陶瓷增殖(HCCB)包层进行热工安全分析。采用大型反应堆瞬态分析程序RELAP5对HCCB包层建模,并进行稳态分析和假设事故的模拟。计算结果表明,CFETR HCCB包层在真空室内氦气泄漏和增殖区盒内氦气泄漏事故中均未出现结构材料熔化,同时各部分的压强变化情况均未超出设计阈值,包层系统在事故发生后均能有效快速地排出余热。CFETR HCCB包层的设计满足热工安全方面的要求。     

10MW高温气冷堆包覆燃料颗粒的研制

我国10MW高温气冷堆采用全陶瓷型包覆颗粒球形燃料元件。TRISO型包覆燃料颗粒由燃料核芯、疏松热解碳层、内致密热解碳层、碳化硅层和外致密热解碳层组成。采用丙烯和乙炔混合气体制备致密热解碳层以及四层连续包覆的新工艺,开展生产工艺条件试验,系统地研究了生产工艺和性能之间的关系,摸索出最佳生产工艺条件。用化学气相沉积方法在150mm流化床沉积炉系统中批量生产出TRISO型包覆燃料颗粒。用扫描电镜观察分析了包覆燃料颗粒的微观结构,包覆燃料颗粒的制造破损率为3.4×10-6,冷态性能达到我国10MW高温气冷堆设计要求。包覆燃料颗粒辐照考验结果(放射性裂变产物释放率R/B为1×10-6左右)表明,包覆燃料颗粒的质量可以满足10MW高温气冷堆安全运行的要求。     

System assessment of an FPGA-based RPS for ABWR nuclear power plant

The instrumentation and control (I&C) systems for the Lungmen nuclear power plant (LMNPP) are fully digitized based on microprocessor and software technology, and extensively utilize multiplexing networks. That is, undetectable software faults and common cause failures due to software errors may occur, and that will defeat the redundancy of a nuclear power plant (NPP). A diverse backup implementation for the digital I&C systems is an important means to defense against undetectable software faults.This paper presents system assessment of a quad-redundant reactor protection system (RPS) design for an Advanced Boiling Water Reactor (ABWR) by utilizing the field programmable gate array (FPGA) technology. The FPGA-based RPS has been assessed by using a full-scope engineering simulator for the LMNPP. Accident scenarios and abnormal conditions are inserted into the engineering simulator in order to activate the function of the FPGA-based RPS. In this study, conceptual design of the proposed quad-redundant FPGA-based RPS, including preliminary hardware architecture, software design and system assessment will be presented. The results demonstrate that the FPGA-based RPS system is a practical approach to implement a diverse backup for the digital I&C system of nuclear power plant applications.Also, the sensitivity study of probabilistic risk assessment (PRA) shows that RPS combined with ARI (Alternative Rod Insertion) contributes significant influence on the core damage frequency (CDF) calculation of LMNPP. The PRA sensitivity study is independent of the RPS technology.     

He冷却试验包层模块的热-力耦合分析

试验包层模块(TBM)是国际热核聚变实验堆(ITER)的关键核心组件,其设计涉及多学科综合优化分析.本文介绍了He冷却固态增殖试验包层的设计概念,并应用热功耦合模拟方法对所提出的包层概念模型的热力响应特性进行分析.结果表明,包层内部各区域的最大温度值和最大应力值均未超过材料容许的限值,所提出的包层设计概念在正常运行工况下是安全可靠的.     

A poloidal field control system based on VxWorks in HT-7 Tokamak

The PF (Poloidal Field) control system is one of the most important control systems in HT-7 Tokamak.Most of parameters such as plasma current, plasma horizontal position and plasma vertical position will be monitored.For the purpose of long-pulse discharge and the more precise control to plasma, the real-time operation system VxWorks is applied, instead of the behindhand and unbefitting DOS operation system. This paper describes the development process of HT-7 PF control system based on VxWorks on Intel X86 platforms. The method of using hardware cards in VxWorks, and the network communication with other operation systems are discussed especially. Results of the comparison between VxWorks and DOS operation systems are given too.     

Application of a 6LiF Small Neutron Detector with an Optical Fiber to Tritium Production Rate Measurement in D-T Neutron Fields

⁶LiF small neutron detectors with an optical fiber have been used to measure ⁶Li(n,α)T reaction rate distributions at thermal research reactors and accelerator facilities. In the present study, we developed an experimental method for the measurement of tritium production rate (TPR) of ⁶Li using this small detector in deuterium-tritium (D-T) neutron fields. Reaction rate measurements with the detector were conducted in the D-T neutron fields at the Fusion Neutronics Source (FNS) facility. From the results, we determined that this detector can be used to measure the TPR distribution in soft neutron spectrum fields such as in a Be assembly. It is difficult to obtain ⁶Li(n,α)T reaction rate separately in hard neutron spectrum fields such as in a Li₂O assembly, because many kinds of charged particle production reactions need to be taken into consideration. However, a time-dependent reaction rate measurement method combined with the ⁶LiF detector and the ZnS detector is effective to separate the ⁶Li(n,α)T reaction from other reactions even in a hard spectrum field, and it can be applied to the measurement of the TPR distribution accurately.     

低杂波电流驱动下HT-7超导托卡马克等离子体硬X射线辐射研究

用７个ＮａＩ（Ｔｌ）闪烁探测器组成的阵列观测能量范围在２０－５００ｋｅＶ,来自ＨＴ－７托卡马克等离子体的硬Ｘ射线轫致辐射。实验结果显示,硬Ｘ射线的径向辐射强度分布与低杂波的功率谱和等离子体的密度及磁志强度等放电参数密切相关。通过测量不同放电参数下硬Ｘ射线的径向强度分布,研究低杂波在等离子体中传播和吸收机制。     

基于FPGA的核安全级仪控系统设计与验证

为解决基于微处理器技术的核电厂安全级数字化仪控系统（DCS）中软件共因故障（CCF）的问题,通过多样性手段避免当未能紧急停堆的预计瞬态（ATWS）发生或反应堆保护系统（RPS）因CCF导致丧失安全功能的风险,本文设计了一种基于现场可编程逻辑门阵列（FPGA）技术的核安全级DCS系统平台,并以核电厂中RPS为实例测试验证平台的功能性能。结果表明:基于FPGA的核安全级DCS系统平台在可用性、适用性和可靠性等方面都满足核电厂安全级数字化仪控系统的要求。      

铀的电子能量损失谱密度泛函理论计算及实验研究

采用密度泛函理论计算及实验获取了铀的电子能量损失谱（Electron energy loss spectroscopy,简称EELS）。计算的谱峰位置与实验一致。结合计算得到铀的能带及态密度对计算谱及实验谱特征峰进行分析,结果表明：由于铀的5f电子形成窄带对等离子体振荡贡献较小,6p电子的共振跃迁致使等离子振荡频率降低;实验谱中13．3eV能量损失峰为体等离子体振荡峰,20．3eV能量损失峰为6p能带到费米能级跃迁能量损失峰,27．6eV能量损失峰为两次体等离子振荡吸收峰。