Use of MCAM in creating 3D neutronics model for ITER building

The three dimensional (3D) neutronics reference model of International Thermonuclear Experimental Reactor (ITER) only defines the tokamak machine and extends to the bio-shield. In order to meet further 3D neutronics analysis needs, it is necessary to create a 3D reference model of the ITER building. Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM) was developed as a computer aided design (CAD) based bi-directional interface program between general CAD systems and Monte Carlo radiation transport simulation codes. With the help of MCAM version 4.8, the 3D neutronics model of ITER building was created based on the engineering CAD model. The calculation of the neutron flux map in ITER building during operation showed the correctness and usability of the model. This model is the first detailed ITER building 3D neutronics model and it will be made available to all international organization collaborators as a reference model.     

Shutdown dose rate calculation code system and its application to EAST tokamak

For safe operation and maintenance of nuclear devices, it is very important to predict the dose rate distribution after shutdown. Based on the rigorous two-step (R2S) method, a new shutdown dose rate calculation code system was developed for nuclear devices with large dimension and complex geometries. By coupling the Monte Carlo particle transport calculation code MCNP with the activation simulation code FISPACT, the dose rate calculation codes integrate the functions of neutron transport calculation, activation calculation and decay gamma transport calculation. This code system was applied to the shutdown dose analysis for experimental advanced superconducting tokamak (EAST). Three-dimensional dose rate distributions of the EAST for different cooling times and spatial locations were obtained. The results can be used to assist radiation protection in EAST.     

ATLAS: Accelerated torus-like angular source for fusion neutronics applications

There is increasing demand for fast and accurate neutronics analysis to be conducted on specific ITER components. The ATLAS (accelerated torus-like angular source) algorithm has been developed to accelerate such neutronics calculations by approximating the 3-D plasma volume source as a surface source at the first-wall region of interest. The ATLAS routine rapidly calculates the angular distribution of the unscattered neutron flux at the chosen first wall location, on a discrete angular grid, and writes a surface source for both MCNP and Attila. This discretised source can then be used to accurately reproduce the full 3-D toroidal volume source in angle and energy, and to quickly simulate the irradiation of individual components or regions in the ITER model. While designed for ITER, the method is general in implementation and can be applied to any tokamak.     

Application of a BC501A Liquid Scintillation Detector with a Gain Stabilization System on the EAST Tokamak

A 2"×2"BC501A liquid scintillation detector with a gain stabilization system is developed and applied to neutron andγ-ray measurement on the EAST tokamak.Energy calibration of a liquid scintillator using a fast coincidence method is presented and compared with the Monte Carlo simulation.Determination of the proton light output function of the BC501A is presented.Results from dedicated experiments with an Am-Be neutron source,γsource and quasi-monoenergetic neutron beams,and from measurements on EAST tokamak are presented and discussed.     

SN方法自动建模程序SNAM在EAST纵场线圈核热计算中的应用

在核领域的中子学计算中,离散纵标(SN)方法三维粒子输运计算程序得到了广泛应用.SNAM作为离散纵标法粒子输运自动建模程序系统,其主要功能是CAD模型与SN计算模型之间的相互转换.一方面,SNAM可以将通用格式的CAD模型转换成SN计算模型,另一方面可以以CAD模型的方式显示SN计算模型中几何及材料等相关信息.本文主要介绍使用SNAM程序对EAST三维模型进行中子学建模,并计算分析了EAST装置纵场线圈的中子注量率与核热沉积.     

托卡马克聚变中子源的数值模型研究

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。     

Internal Magnetic Configuration Measured by ECE Imaging on EAST Tokamak

ECE imaging （electron cyclotron emission imaging） is an important diagnostic which can give 2D imaging of temperature fluctuation in the core of tokamak. A method based on ECE imaging is introduced which can give the information of the position of magnetic axis and the structure of internal magnetic surface for EAST tokamak. The EFIT equilibrium reconstruction is not reliable due to the absence of important core diagnostic at the initial phase for EAST, so the information given by ECE imaging could help to improve the accuracy of EFIT equilibrium reconstruction.     

核聚变实验装置HT-7U一维及二维辐射防护设计研究

主要介绍一维、二维中子输运程序ANISN,DOT3.5在核聚变实验装置HT-7U辐射屏蔽物理设计中的应用。计算和分析了该装置实验大厅内外中子注量/能谱、γ注量/能谱、中子剂量率、γ剂量率的空间分布,对屏蔽材料的选取及屏蔽层厚度进行了优化设计,为HT-7U装置的辐射屏蔽物理设计提供了建议性意见及理论依据。     

蒙特卡罗模拟CFETR中子输运计算中的全局减方差方法应用及对比

在具有全局特性的蒙特卡罗输运精细计算的问题中,传统的MCNP(Monte Carlo N Particle Transport Code)局部减方差方法很难得到理想的计算结果,全局减方差方法(Global Variance Reduction,GVR)则是一种有效的解决方法。针对中国聚变工程试验反应堆(Chinese Fusion Engineering Testing Reactor,CFETR)的中子输运过程中减小全局方差的问题,将多种形式的GVR方法应用到柱状CFETR中子学模型的计算中。依据不同的中子分布信息,在算例中应用和对比了6种不同形式的GVR权窗,并对不同GVR方法的品质因子(FOMG)、标准差(σ)和有效计数率(Scoring)进行了分析。与AN(MCNPanalog method)相比,GVR方法的FOMG有很大的增长,误差在空间的分布也更加平缓,且具有更高的Scoring。在前人提出的全局减方差的基础上,在计算中应用一些新的GVR形式(能量、径迹数等),计算结果表明,基于中子通量的GVR方法的全局计算效率较AN提高了6.43倍。此外,基于中子能量的全局减方差方法也是一种可行的GVR应用形式,其与AN比较,计算效率提高了5.11倍。综上,基于中子通量的GVR方法具有最佳的全局减方差效果。     

The ITER radial neutron camera: An updated neutronic analysis

The radial neutron camera (RNC) will provide the spatial distribution and the total strength of the ITER neutron source (emissivity profile and fusion power) by means of collimated neutron measurements. Line-integrated neutron spectral measurements can also provide information on the ion temperature profile. The present design of the RNC consists of two collimating structures for a full coverage of the plasma: 36 collimated lines of sight (LOS) distributed in three different planes view the plasma core (ex-port system) and nine collimated LOS view the plasma edge (in-port system).The RNC design is based on the combined use of the MCNP Monte Carlo code and a software tool performing asymmetric Abel inversion of simulated measured neutron signals (MSST). Neutron and γ-ray transport calculations are performed with MCNP using a 3D RNC model to determine the signal/noise for each RNC channel and the spectra at the detectors. The MSST code is used to check the RNC compliance with the ITER measurement requirements for the neutron emissivity profile.In the present paper the improvement of the hard variance reduction technique applied to the MCNP neutron source (consisting in sampling neutrons only from plasma regions contributing to the detector signal) is presented and the following issues are analyzed: the possibility of reducing the length of the ex-port collimators (resulting in a significant reduction of the overall RNC dimension and weight); options for the reduction of the dose due to the neutron streaming through the RNC cut-outs in the blanket shielding module; the integration of a γ-ray detection system in the RNC by partially filling the collimators with a neutron absorbing material (LiH).     

Calculation of Neutral Beam Injected Torque and Its Effective Tangency Major Radius for EAST

Toroidal rotation has been recognized to have significant effects on the transport and magnetohydrodynarnic(MHD) stability of tokamak plasmas.Neutral beam injection(NBI) is the most effective rotation generation method on current,tokamak devices.To estimate the effective injected torque of the first neutral beam injection system on EAST,a simplified analytic method was derived.Calculated beam torque values were validated by those obtained from the NUBEAM code simulation.According to the results,for the collisional torque,the effective tangential radius for torque deposition is close to the beam tangency major radius.However,due to the dielectric property of tokamak plasma,the equivalent tangency major radius of the J×B torque is equal to the average major radius of the magnetic flux surface.The results will be useful for the research of toroidal momentum confinement and the experimental analysis of momentum transport related with NBI on EAST.     

快热耦合系统的中子学计算工具——采用确定论方法

对于既具有高能量外中子源又包含热中子区域的快热耦合系统，本文将离散坐标法Sn与栅格计算程序相结合，给出可用于快热耦合系统的中子学确定论计算工具。用该方法计算了启明星1号次临界装置的k_eff，并给出了计算结果。     

Neutron yield measurement system of HL-2A tokamak

This research presents the development of HL-2A neutron yield measurement which includes ²³⁵U fission chamber and BF3 and ³He proportional counters. Equivalent noise formula of the radiation detection signal amplification system was derived to guide the development of the signal amplification system. Then all detectors were calibrated in situ by using the ²⁵²C_f neutron source. The neutron yield of the HL-2A during neutral beam heating was analyzed. These results indicate that the developed neutron flux diagnostic system can obtain neutron yield results under various experimental conditions of the HL-2A tokamak, and can provide information on neutron yield.     

泄漏源首次碰撞补偿技术

提出泄漏源首次碰撞补偿技术,以解决二维/一维中子输运计算收敛不稳定问题。将源项通过首次碰撞方法等效为各个区域的散射源,相当于将局部的孤立源分布到整体的广泛空间中,从而减轻泄漏源加重的射线效应影响,并应用单能修正简化计算方法,提高了二维/一维中子输运计算的收敛性、稳定性和精度。      

Effect of Fusion Neutron Source Numerical Models on Neutron Wall Loading in a D—D Tokamak Device

Effect of various spatial and energy distributions of fusion neutron source on the calculation of neutron wall loading of Tokamak D-D fusion device has been investigated by means of the 3-D Monte Carlo code MCNP.A realistic Monte Carlo source model was developed based on the accurate representation of the spatial distribution and energy spectrum of fusion neutrons to solve the complicated problem of tokamak fusion neutron source modelling.The results show that those simplified source models will introduce significant uncertainties.For accurate estimation of the key nuclear responses of the tokamak design and analyses,the use of the realistic source is recommended.In addition,the accumulation of tritium produced during D-D plasma operation should be carefully considered.     

Synthesis of Li2O Single Crystal by Vacuum Fusion Technique

Detailed neutronics analysis on a tokamak fusion experimental reactor was performed. Neutronics responses of components in the vacuum vessel of a swimming pool type reactor were calculated by Monte Carlo method considering toroidal geometry, asymmetrical torus cross section, and neutron source distribution in plasma. It was found that the neutronics responses in the first wall and the blanket vary significantly in the poloidal direction. Effect of penetrations of neutral beam injectors on the tritium breeding performance was also evaluated in the analysis. The result shows that the effect is rather small in the present structure. As a result of the analysis, it becomes clear that modification of blanket structure and material composition should be done in order to improve the tritium breeding performance and to reduce the nuclear heating rate of the vacuum vessel in the divertor zone.     

托卡马克实验装置的实验大厅移动辐射监测机器人设计

托卡马克实验装置(experimental advanced superconducting tokamak, EAST)在进行聚变等离子体放电实验时,随着中性束注入、低杂波电流驱动、电子回旋波共振加热等大功率辅助加热系统投入运行,等离子体中子出射强度可达10¹⁴ n/s,这导致大厅及窗口区域的中子光子剂量率明显增加。为了保证工作人员及维护人员的安全,需要及时了解不同位置处的辐射剂量。本研究基于六轮移动救援机器人底盘,研制EAST实验大厅移动辐射监测机器人,并搭载X、γ辐射在线监测仪,通过同步定位与建图技术（SLAM）构建环境地图自主移动或通过高清图像远程控制及数据传输系统,将大厅内不同位置的辐射剂量值传递给上位机。通过巡测路径上的剂量率异常,可向周围的工作人员及时发出警示,以此确保实验人员的辐射安全和判断现场维护人员的滞留时间。     

Calculation of the Heat Flux on the First Wall During Disruption on Tokamak

Disruptions are the most dangerous instabilities in tokamak plasma. During plasma disruption, the large amounts of energy will be deposited on Plasma Facing Components (PFCs) which is a damaging threat for the divertor target and the first wall materials. Therefore, studying the characteristic of heat deposition on the first wall is very significant. The Infrared (IR) camera is an effective tool to measure the surface temperature profile on the first wall on the Experimental Advanced Superconducting Tokamak (EAST). With a finite difference method, the heat flux arrived to the divertor can be calculated from the surface temperature. However, the surface layer on the divertor has a great influence on the calculation of the heat flux on the divertor. The numerical method for solving heat conduction for semi-infinite model is given in this paper. And the thermal resistance of surface layers is considered in this numerical method. In addition, the distribution of heat flux on the divertor during disruption is also shown.     

Assessment of Radiation Damage to the Structural Material of EAST Tokamak

Radiation damage to structural material of fusion facilities is of high concern for safety. The superconducting tokamak EAST will conduct D-D plasma experiments with the neutron production of 1015 neutrons per second. To evaluate the material radiation damage a programme system has been devised with the Monte Carlo transport code MCNP-4C, the inventory code FISPACT99, a specific interface, and the fusion evaluated nuclear data library FENDL-2. The key nuclear responses, i.e. fast neutron flux, displacement per atom, and the helium and hydrogen production, are calculated for the structural material SS-316L of the first wall, and the vacuum vessel, using this programme. The results demonstrate that the radiation damage to the structural material is so little that it will not lead to any significant change of material properties according to the reference design. This indicates that there is a large potential space for EAST to test advanced operation regime from the viewpoint of structural material safety.