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Simulation of reactor pulses in fast burst and externally driven nuclear assemblies.

机译:快速爆炸和外部驱动核组件中反应堆脉冲的仿真。

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摘要

The following research contributes original concepts to the fields of deterministic neutron transport modeling and reactor power excursion simulation. A deterministic neutron transport code was created to assess the value of new methods of determining neutron current, fluence, and flux values through the use of view factor and average path length calculations. The neutron transport code is also capable of modeling the highly anisotropic neutron transport of deuterium-tritium fusion external source neutrons using diffusion theory with the aid of a modified first collision source term. The neutron transport code was benchmarked with MCNP, an industry standard stochastic neutron transport code.;Deterministic neutron transport methods allow users to model large quantities of neutrons without simulating their interactions individually. Subsequently, deterministic methods allow users to more easily couple neutron transport simulations with other physics simulations. Heat transfer and thermoelastic mechanics physics simulation modules were each developed and benchmarked using COMSOL, a commercial heat transfer and mechanics simulation software. The physics simulation modules were then coupled and used to simulate reactor pulses in fast burst and externally driven nuclear assemblies.;The coupled system of equations represents a new method of simulating reactor pulses that allows users to more fully characterize pulsed assemblies. Unlike older methods of reactor pulse simulation, the method presented in this research does not require data from the operational reactor in order to simulate its behavior. The ability to simulate the coupled neutron transport and thermo-mechanical feedback present in pulsed reactors prior their construction would significantly enhance the quality of pulsed reactor pre-construction safety analysis. Additionally, a graphical user interface is created to allow users to run simulations and visualize the results using the coupled physics simulation modules. (Copies available exclusively from MIT Libraries, Rm. 14-0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253-1690.)
机译:以下研究为确定性中子输运建模和反应堆功率偏移模拟领域提供了原始概念。创建了确定性的中子输运代码,以通过使用视场因子和平均路径长度计算来评估确定中子电流,通量和通量值的新方法的价值。中子输运代码还能够利用扩散理论借助于修改的第一碰撞源项,对氘-fusion聚变外部源中子的高各向异性中子输运进行建模。中子传输代码以MCNP(行业标准随机中子传输代码)为基准。确定性中子传输方法使用户可以对大量中子建模,而无需单独模拟它们的相互作用。随后,确定性方法使用户可以更轻松地将中子输运模拟与其他物理模拟耦合。传热和热弹性力学物理模拟模块均使用COMSOL(商业传热和力学模拟软件)进行开发和基准测试。然后耦合物理仿真模块,并用于仿真快速爆炸和外部驱动的核组件中的反应堆脉冲。方程组耦合表示一种模拟反应堆脉冲的新方法,使用户可以更全面地表征脉冲组件。与旧的反应堆脉冲模拟方法不同,本研究提出的方法不需要运行中的反应堆数据即可模拟其行为。在脉冲反应堆建造之前模拟脉冲中子耦合的中子输运和热机械反馈的能力将显着提高脉冲反应堆施工前安全性分析的质量。另外,创建了图形用户界面,以允许用户使用耦合的物理模拟模块运行模拟并可视化结果。 (仅可从麻省理工学院图书馆14-0551室,剑桥,马萨诸塞州02139-4307;电话617-253-5668;传真617-253-1690获得副本。)

著录项

  • 作者

    Green, Taylor Caldwell, IV.;

  • 作者单位

    The University of Texas at Austin.;

  • 授予单位 The University of Texas at Austin.;
  • 学科 Engineering Mechanical.;Engineering Nuclear.
  • 学位 Ph.D.
  • 年度 2008
  • 页码 188 p.
  • 总页数 188
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类 机械、仪表工业;原子能技术;
  • 关键词

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