Electromagnetic and Structural Analysis on Vacuum Vessel for CFETR During Plasma Major Disruption

China Fusion Engineering Test Reactor (CFETR) is a superconducting magnet tokamak and its goal is to achieve the magnetic confinement fusion. The electromagnetic (EM) transients cause mechanical forces, which represent one of the most vital loads for tokamak vacuum vessel (VV). This paper is focused on calculational methods and results for the EM loads on the simplified but practical model of CFETR VV with respect to plasma major disruption scenarios as a reference of the design and analysis. Commercial finite element method software, ANSYS, was employed to evaluate the eddy current on the VV module with the 22.5 ° sector model for major conducting structure of the tokamak including double-walled VV, T-shape rib, and three ports. The plasma current is damping as exponential function 36 ms corresponding to the current simulating in ITER outputs, which are one of major sources of EM loads on VV components. As the results of calculating the eddy currents and EM forces, stress and deformation on CFETR VV can be obtained, which is useful for the structural design of VV.     

cosRMC复杂曲面开发及在聚变中子学的应用

中国聚变工程实验堆（CFETR）是我国自主设计和研制的重大科学工程,CFETR旨在与ITER相衔接和补充,为研制DEMO级别聚变堆电站提供必要的技术。蒙特卡罗方法在聚变中子学与屏蔽设计等方面具有重要作用。本文基于自主化蒙特卡罗程序cosRMC,研究了蒙特卡罗复杂曲面建模的数学模型和计算方法,开发了复杂曲面建模功能,并通过PPCS（power plant conceptual study）模型验证了该功能实现的正确性。然后构建了CFETR的三维精细化模型,并利用该模型对CFETR包层设计中的关键中子学参数进行计算分析。结果表明,cosRMC对中子学参数氚增殖比、中子壁载荷和核热沉积的计算结果与MCNP的计算值吻合良好,相对偏差均小于5%,满足工程设计需求。研究证明了cosRMC应用于聚变堆包层中子学分析的正确性和有效性。CFETR中子学参数的计算分析,也为其设计和优化提供了参考。     

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。     

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.     

Impact Analysis of Helium Cooled Solid Blanket Structures on the Tritium Breeding Performance in CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant. As its objectives are to demonstrate generation of fusion power and to realize tritium self-sufficiency, the tritium breeding ratio (TBR) is a key design parameter. In the blanket design and optimization, the structures such as the first wall (FW), cooling plate (CP), stiffening plate (SP), cap and some other design parameters in detailed 3-D model have significant impacts on the tritium breeding performance. Based on a helium cooled solid breeder blanket option for CFETR, the impact analysis of the helium cooled solid blanket structures on tritium breeding performance was performed in this paper. Firstly, the detailed 3D neutronics model was built by using of a CAD to Monte Carlo Geometry conversion tool McCad. Then based on the detailed 3D neutronics model, the impact analyses of the blanket structures on tritium breeding performance were carried out, which include the FW, CP, SP, cap and side wall. By the sensitivity study of the blanket structures on the TBR, it gave the TBR variation trend and references for the blanket design and optimization.     

A study on nuclear analysis of the divertor region of the CFETR

This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor(CFETR)divertor region using the MCNP-5 Monte Carlo N-particles code in a 3D geometry model.We assessed the nuclear responses of the divertor region component systems and evaluated their shielding capability,which can support the development strategy of the physical and engineering design of the CFETR.Model specification based on the latest CAD model of the CFETR divertor has been integrated into the CFETR MCNP reference model with a major/minor radius R=7.2 m/a=2.2 m in the 22.5° model,and a fusion-power range of around 1-1.5 GW.The nuclear heating and radiation damage of the divertor system are enhanced compared to that of the ITER and the earlier CFETR design.The initial nuclear responses of the toroidal field coil and vacuum vessel systems showed that the shielding of the current divertor design is not sufficient and optimization work has been carried out.We also carried out calculations and analysis using a hypothetical operating scenario of over 14 years.An excellent improvement in the nuclear performance has been obtained by the improved additional shielding block in the divertor region when referring to the ITER design limit,which can support the design of the future update of the divertor region systems of the CFETR.     

Proposal of Integrated Test Facility for in-Vessel Thermofluid Safety of Fusion Reactors

A vacuum vessel (VV) of a tokamak fusion reactor like the International Thermonuclear Experimental Reactor (ITER) consists the first confinement barrier that includes the largest amount of radioactive materials such as tritium and activation products. The ingress of coolant event (ICE) is a design basis event in the ITER where water is used as the coolant. The loss of vacuum event (LOVA) is also considered as an independent design basis event. Based on the results of ICE and LOVA preliminary experiments, an integrated in-vessel thermofluid test is being planned and conceptual design of the facility is in progress. The main objectives of the integrated test are to investigate the consequences of possible interaction of the ICE and the LOVA and to validate the analytical model of thermofluid events in the VV of the fusion reactor. This paper introduces a conceptual design of the integrated test facility and a testing plan.     

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。     

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

China Fusion Engineering Test Reactor(CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO.One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2to ensure tritium self-sufficiency.A concept design for a water cooled ceramics breeding blanket(WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR.Based on this concept,a one-dimensional(1D) radial built breeding blanket was first designed,and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build.A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models,addressing neutron wall loading(NWL),tritium breeding ratio(TBR),fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components.The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design.     

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。     

Concept design of CFETR superconducting magnet system based on different maintenance ports

CFETR which stands for “China Fusion Engineering Test Reactor” is a new tokamak device. Its magnet system includes the Toroidal Field (TF) winding, Center solenoid winding (CS) and Poloidal Field (PF) winding. The main goal of the project is to build a fusion engineering Tokamak reactor with its fusion power is 50–200 MW and should be self-sufficiency by blanket.In order to ensure the maintenance ports design and maintenance method, this article discussed the concept design of the magnet system based on different maintenance port cases. The paper detailed studied the magnet system of CFETR including the electromagnetic analysis and parameters for TF (CS)PF. Besides, the volt-seconds of ohmic field are presented as detailed as possible in this paper. In addition, the calculations and optimizations of equilibrium field which should guarantee the plasma discharge of single null shape is carried out. The design work reported here illustrates that the present maintenance ports will not have a great impact on the design of the magnet system. The concept design of the magnet system can meet the requirement of the physical target.     

Effects of kinetic profiles on neutron wall loading distribution in CFETR

The China Fusion Engineering Test Reactor (CFETR) is under design,which aims to bridge the gaps between ITER and the future fusion power plant.The neutron wall loading (NWL) depends on the neutron source distribution,which depends on the density and temperature profiles.In this paper,we calculate the NWL of CFETR and study the effects of density and temperature profiles on the NWL distribution along the first wall.Our calculations show that for a 200 MW fusion power,the maximum NWL is at the outer midplane and the vaule is about 0.4 MW m-2.The density and temperature profiles have little effect on the NWL distribution.The value of NWL is determined by the total fusion power.     

An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

The Chinese Fusion Engineering Tokamak Reactor(CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder(WCCB)blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic(RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement.     

Process Design of Cryogenic Distribution System for CFETR CS Model Coil

The superconducting magnet of Central Solenoid(CS) model coil of China Fusion Engineering Test Reactor(CFETR) is made of Nb_3Sn/Nb Ti cable-in-conduit conductor(CICC),and operated by forced-flow cooling with a large amount of supercritical helium.The cryogenic circulation pump is analyzed and considered to be effective in achieving the supercritical helium(SHe) circulation for the forced-flow cooled(FFC) CICC magnet.A distributed system will be constructed for cooling the CFETR CS model coil.This paper presents the design of FFC process for the CFETR CS model coil.The equipment configuration,quench protection in the magnet and the process control are presented.     

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。     

Design,Analysis and Optimization of the First Wall Cooling System of the CFETR Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak which is designed by China National Integration design Group for Magnetic Confinement Fusion. CFETR Blanket, as a plasma-facing component withstand very high heat load, is very critical for fusion reactor operation. The first wall (FW) is one of the most significant components of the blanket. The cooling system of the FW has been designed. Meanwhile, thermal–dynamic calculations are performed to obtain the coolant feature and temperature distribution of the FW using ANSYS CFX code. Besides, thermo-mechanical coupling analysis is carried out using the temperature distribution from thermal–dynamic calculation as boundary condition. In addition, cooling channel optimization is proposed according to the analysis results. Analysis results of the optimization cooling channel indicate that the maximum temperature and thermal stress satisfy the design requirements of the FW.     

RAMI analysis of the ITER CIS

ITER is the first worldwide international project aiming to design a device that proves the physics and technological basis for fusion power plants to produce nuclear fusion energy. In the project, the RAMI approach (reliability, availability, maintainability and inspectability) has been adopted for technical risk control to guide the design of components in preparation for operation and maintenance. RAMI analysis of the ITER central interlock system (CIS), which shall provide investment protection for the ITER systems was performed on the conceptual design. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 5 main functions and 7 sub-functions which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to estimate the reliability and availability of each function under stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CIS expected after implementation of mitigating actions was calculated to be 99.86% over 2 years, which is the typical interval of the scheduled maintenance cycles. A failure modes, effects and criticality analysis (FMECA) was performed to initiate risk mitigation action. Criticality matrices highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. It was assessed that the availability of the ITER CIS, with appropriate mitigating actions applied, meets the project availability requirement for the system.     

RAMI Analysis of HCCB TBS for ITER

ITER is the first worldwide international experimental nuclear fusion facility, which aims to prove the physics and technological basis for future fusion power plants. As main stages of ITER technical risk control, the reliability, availability, maintainability and inspectability (RAMI) approach should be applied to all ITER components during their design phase to reduce potential technical risks. Test blanket modules play a key role in ITER. Helium cooled ceramic breeder (HCCB) TBM is one of TBM concepts which were proposed by China. HCCB TBM and its ancillary system are called HCCB test blanket system (TBS). The RAMI analysis was performed on the conceptual design of the ITER HCCB TBS in this paper. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 3 main functions, 1 support function, 14 sub-functions and 50 basic functions. These functions were described using the IDEF0 method. Reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the HCCB TBS expected after implementation of mitigation actions was calculated to be 94.69 % over 2 years. A failure modes, effects and criticality analysis was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.     

COSINE堆用蒙卡分析软件cosRMC研发与应用

为进一步提升核电软件自主化能力,研发了核电厂设计与安全分析一体化软件包COSINE。其中cosRMC为堆用三维中子-光子-电子输运蒙卡软件,已具备输运计算、燃耗计算、群常数产生、敏感性及不确定性分析、可视化建模等功能,可用于堆芯设计分析、确定论校核计算以及辐射屏蔽计算。本文从cosRMC的计算功能以及软件在先进非能动型压水堆(AP1000)与中国聚变工程实验堆(CFETR)中的典型应用对cosRMC软件的研发现状进行介绍。其中,AP1000堆芯的模拟结果显示,21种燃料组件及全堆芯模型的增殖因子绝对值最大偏差为89.9×10~(-5),功率分布计算结果绝对值最大偏差为2.1%;CFETR的模拟结果显示,氚增殖比的最大绝对值偏差为0.6%,cosRMC网格权窗功能可以有效解决模拟过程中的深穿透问题。cosRMC软件计算功能可满足压水堆、聚变堆等大型复杂模型的计算需求,软件具有较高的计算精度,同时可视化建模工具可有效提升建模效率及正确性。