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In the previous chapter, we have assumed that the neutron source is given. This is not the case in the core of a nuclear reactor, since the fission source depends on the neutron flux. Mathematically, we obtain an eigenvalue problem that we shall be able to solve in a few particular cases. From such an eigenvalue problem, we shall be able to deduce a critical equation, at least in the homogeneous case.

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One Group Diffusion Model

  • 章节

  • 发表日期：

    2025年

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  • Bertrand Mercier

摘要

In the previous chapter, we have assumed that the neutron source is given. This is not the case in the core of a nuclear reactor, since the fission source depends on the neutron flux. Mathematically, we obtain an eigenvalue problem that we shall be able to solve in a few particular cases. From such an eigenvalue problem, we shall be able to deduce a critical equation, at least in the homogeneous case.