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Numerical methods for nuclear fuel burnup calculations

机译:核燃料燃耗计算的数值方法

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摘要

The material composition of nuclear fuel changes constantly due to nuclides transforming to other nuclides via neutron-induced transmutation reactions and spontaneous radioactive decay. The objective of burnup calculations is to simulate these changes over time. This thesis considers two essential topics of burnup calculations: the numerical solution of burnup equations based on computing the burnup matrix exponential, and the uncertainty analysis of neutron transport criticality equation based on perturbation theory.The burnup equations govern the changes in nuclide concentrations over time. They form a system of first order differential equations that can be formally solved by computing the matrix exponential of the burnup matrix. Due to the dramatic variation in the half-lives of different nuclides, the system is extremely stiff and the problem is complicated by vast variations in the time steps used in burnup calculations. In this thesis, the mathematical properties of burnup matrices are studied. It is deduced that their eigenvalues are generally confined to a region near the negative real axis. Rational approximations that are accurate near the negative real axis, and the Chebyshev rational approximation method (CRAM) in particular, are proposed as a novel method for solving the burnup equations. The results suggest that the proposed approach is capable of providing a robust and accurate solution to the burnup equations with a very short computation time.When a mathematical model contains uncertain parameters, this uncertainty is propagated to responses dependent on the model. This thesis studies the propagation of neutron interaction data uncertainty through the criticality equation on a fuel assembly level. The considered approach is based on perturbation theory, which allows computing the sensitivity profiles of a response with respect to any number of parameters in an efficient manner by solving an adjoint system in addition to the original forward problem. The uncertainty related to these parameters can then be propagated deterministically to the response by linearizing the response.
机译:核燃料的物质组成由于核素通过中子诱发的mut变反应和自发放射性衰变而转化为其他核素而不断变化。燃耗计算的目的是模拟这些随时间的变化。本文考虑了燃耗计算的两个基本主题:燃耗方程的数值解基于燃耗矩阵指数的计算,以及基于微扰理论的中子输运临界方程的不确定性分析。燃耗方程控制着核素浓度随时间的变化。它们形成了一阶微分方程组,可以通过计算燃耗矩阵的矩阵指数来正式求解。由于不同核素半衰期的剧烈变化,该系统非常僵硬,并且燃耗计算中所用时间步长的巨大变化使问题变得复杂。本文研究了燃耗矩阵的数学性质。可以推断,它们的特征值通常被限制在负实轴附近。提出了一种在负实轴附近精确的有理逼近,尤其是Chebyshev有理逼近方法(CRAM),作为求解燃耗方程的一种新方法。结果表明,该方法能够以极短的计算时间为燃耗方程提供鲁棒且准确的解决方案。当数学模型包含不确定参数时,此不确定性会传播到依赖于模型的响应中。本文通过在燃料组件水平上的临界方程研究中子相互作用数据不确定性的传播。所考虑的方法基于微扰理论,该理论允许通过求解伴随系统以及原始正向问题,以有效的方式针对任何数量的参数计算响应的灵敏度曲线。然后可以通过线性化响应,将与这些参数有关的不确定性确定性地传播到响应中。

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    Pusa Maria;

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