共查询到20条相似文献,搜索用时 234 毫秒
1.
聚变驱动次临界堆双冷嬗变包层材料活化计算与分析 总被引:1,自引:1,他引:0
对聚变驱动次临界堆 (FDS Ⅰ )包层进行了材料活化计算与分析。利用多功能中子学程序系统VisualBUS1 .0及多群数据库HENDL1 .0 /MG进行中子输运计算 ,以获得包层各个功能区的中子注量率能谱 ;在此基础上 ,使用欧洲活化计算程序FISPACT及IAEA聚变活化数据库FENDL/A 2 .0分别对停堆初期包层不同功能区的剂量率水平和衰变余热水平、停堆后期结构材料与氚增殖剂 /冷却剂的活化性能及其杂质的控制要求进行了计算及分析。 相似文献
2.
为了验证双功能锂铅实验包层模块(DFLL-TBM)的中子学设计,中国科学院核能安全技术研究所·FDS凤麟核能团队利用14 MeV中子源开展了DFLL-TBM模型的中子学实验。实验中分别利用In、Al、Nb活化片和~6Li玻璃探测器测量了DFLL-TBM中子学实验模型中不同深度3个位置的活化反应率和产氚率。并利用蒙特卡洛模拟程序Super MC和FENDL3.1数据库进行了相应的模拟计算,计算值和实验值比较在10%以内吻合。结果表明计算值与实验值符合较好,所采用的计算程序和数据库适用于DFLL-TBM的计算设计。 相似文献
3.
为验证在中国先进研究堆(CARR)内进行国际热核聚变实验堆(ITER)氚增殖包层模块(TBM)辐照实验的可行性和安全性,进行了氚增殖剂球床组件堆内辐照物理及热工计算分析。氚增殖剂包层模块主要是固态氚增殖剂陶瓷球床。本文采用Monte Carlo粒子输运模拟程序对氚增殖剂球床进行堆内建模,计算球床的中子注量率、能量沉积和产额,得到不同功率下球床的中子注量率、发热功率和产氚速率以及球床组件引入反应堆的反应性。根据物理计算得到的组件各部件发热情况建立热工计算一维模型,通过更改反应堆功率得到满足实验要求的工况并采用三维程序进行验证。物理与热工计算分析的结果表明,在反应堆运行功率为20 MW的工况下球床组件各部件的温度均不超过限值。 相似文献
4.
5.
反应堆压力容器(RPV)中的碳钢材料受到快中子辐照会发生性能变化。为了防止由于RPV的材料性能发生变化而不适当地限制核电厂的运行,需要限定核电厂寿期内RPV中的最大快中子注量,并且要求安装辐照监督管对RPV材料所受到的快中子注量进行监督。因此,RPV和辐照监督管中子注量率的精确计算对RPV的辐照安全和寿命管理具有十分重要的意义。三代非能动压水堆核电厂主要采用基于BUGLE-96截面库的二维离散纵标法程序DORT进行RPV中子注量率计算。本文利用秦山核电厂第五根辐照监督管的中子注量率测量数据和MCNP-4B计算结果与DORT程序的计算结果进行比较,来验证采用DORT程序进行RPV母材段中子注量率计算的可靠性。 相似文献
6.
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。 相似文献
7.
辐射屏蔽计算是核电厂辐射防护设计和审评的重要内容之一。国际屏蔽计算软件对中国实行"出口封锁",制约了我国核电辐射屏蔽审核计算能力,因此,研发了具有自主知识产权的基于蒙特卡罗方法的辐射屏蔽专用蒙特卡罗软件RShieldMC(Radiation Shielding Monte Carlo)。为了验证RShieldMC程序,进行中子注量率计算的准确性和适用性,利用秦山一期反应堆结构与辐照监督管相关参数,通过RShieldMC可视化前处理模块建立辐照监督管屏蔽计算模型,计算秦山一期反应堆辐照监督管堆芯中平面和上焊缝处的中子注量率。RShieldMC程序计算结果与辐照监督管实验测量值以及MCNP(Monte Carlo N-Particle)、JMCT-S(J Monte Carlo Transport)、TORT(Three-dimensional Neutron/Photon Transport)程序计算结果符合较好,验证了RShieldMC软件在中子注量率计算中的可用性及正确性。 相似文献
8.
9.
10.
11.
Guangchun Zhang Hongchun Wu Liangzhi Cao Youqi Zheng Yunzhao Li Zhouyu Liu 《Fusion Engineering and Design》2013,88(5):413-420
The neutronics analysis on the test blanket module (TBM) has important significance for the ITER device and its related experiment design. Quantities of scoping-type studies and conceptual designs were published by using the Monte Carlo method. However, disadvantages like time consuming make it necessary to develop a new highly efficient method. Hence, a new two-step approach method based on the 3D deterministic method for analyzing the TBM is proposed in this paper. A code package 3DMOC-NSPn was developed. It is mainly composed of three modules, the 3DMOC for generating the homogenization cross section; the LINK code for cross section condensation and the NSPn code for blanket calculation. The detailed flux distribution throughout the whole TBM and the mainly neutronics features, such as TBR, displacement per atom (DPA), helium and hydrogen production rate can be obtained. To validate the numerical approach and the code package, the calculations on China dual functional lithium lead-test blanket module (DFLL-TBM) was performed. The reference results were obtained by using the MCNP code. The numerical results from 3DMOC-NSPn are in good agreement with the references. It indicates that the whole code package is a reliable neutronics analysis tool for the TBM design and evaluation. 相似文献
12.
为了满足ITER对波纹度的要求,核工业西南物理研究院提出了新的减少低活化铁素体钢的氦冷固态(HCSB)实验包层模块(TBM)设计方案。采用MCNP程序及ITER全堆MCNP模型,对新设计的2×6HCSB-TBM进行三维中子学计算分析,给出了模块产氚率、核热沉积和功率密度分布等结果。在ITER运行因子为22%时,HCSB-TBM的产氚率为12.68mg/d。TBM内总核热沉积为522.5kW,最高功率密度为11.8W/cm3,出现在氚增殖区Li4SiO4中。计算结果可为TBM进一步的结构、热工水力学优化及其他系统设计提供中子学数据。 相似文献
13.
The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR).Some updating of neutronics analyses was needed,because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket,including the optimization of radial build-up and customized structure for each blanket module.A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses.The tritium breeding capability,nuclear heating power,radiation damage,and decay heat were calculated by the MCNP and FISPACT code.The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency.The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW.The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60,respectively.The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module # 3.The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time.The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years. 相似文献
14.
激光惯性约束聚变裂变混合能源包层中子学数值模拟 总被引:1,自引:1,他引:0
对三维输运与燃耗耦合程序MCORGS进行了适应性改造,并对利弗莫尔实验室提出的激光惯性约束聚变裂变混合能源(LIFE)概念进行了分析和改进。输运计算采用MCNP程序,燃耗计算采用ORIGENS程序,增加氚控制模块和功率控制模块。建立了与LIFE等价的以贫化铀为燃料、Be为中子增殖剂的包层方案,通过数值模拟验证了MCORGS程序的可靠性。针对Be资源短缺及冷却复杂问题,设计了以贫化铀为燃料、Pb为中子增殖剂的包层方案,包层能量放大了4倍,可在55a内稳定输出2 000 MWt功率。 相似文献
15.
16.
17.
18.
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. 相似文献
19.
《等离子体科学和技术》2016,18(8):865-869
Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. 相似文献
20.
This paper presents a new 1D Neutronics/Thermal-hydraulics code ATAC-1D based on the advanced Jacobian-Free Newton-Krylov (JFNK) method and the low dimensional equivalent strategy. Conventional operator-splitting (OS) strategies are used to maintain its accuracy with small time steps and linearization of the nonlinear problem, which leads to slow computation speed and linearization error. The JFNK method solves the troubles in the coupled neutronics/thermal-hydraulics problems mentioned above. Furthermore, a core-wide three dimension to one dimension equivalent method has been developed to provide variable few-group parameters. Finally, the performance of the coupled neutronics/thermal-hydraulics code ATAC-1D is studied by simulating four OECD/NEA CRP PWR rod ejection benchmark problems. The simulation results are compared to the reference ones, which proves that the developed 1D code has a good accuracy and practicability in nuclear reactor transient calculation. 相似文献