Solving the neutron slowing down equation Bertrand Mercier, Jinghan Peng
To cite this version:
Bertrand Mercier, Jinghan Peng. Solving the neutron slowing down equation. 2014. hal-01081772
HAL Id: hal-01081772 https://hal.archives-ouvertes.fr/hal-01081772 Preprint submitted on 11 Nov 2014
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Distributed under a Creative Commons Attribution - NonCommercial| 4.0 International License Bertrand Mercier - Arthur Peng
Solving the neutron slowing down equation.
Abstract : It is well-known that the neutron slowing down equation has an exact solution in the case where the moderator is only made of hydrogen (A=1). In this paper we consider the case where the moderator is made with heavy water (A=2) and show that it is possible to approximate the transition probability by a law with the same expectation, but leads to an analytic solution of the same type as for A=1. We also show that such an analytic solution gives a lot of information about what happens in the core of a nuclear reactor, and gives a way to compute self-shielded cross sections.
Introduction : Fission neutrons are fast neutrons (energy in the range 1 to 10 MeV).
However, Uranium 235 fission cross section is much higher ( σf =580 barns) for thermal neutrons
(energy smaller than 1 eV) than for fast neutrons (σf = 2 barns). This basic feature is used in most nuclear reactors to reduce the size of the core. A moderator (see e.g. http://en.wikipedia.org/wiki/Neutron_moderator) is then needed to achieve neutron slowing down. Neutron slowing down in a reactor is a complex phenomenon due to the fact that nuclear fuel is usually made of a combination of 238 U and 235 U, with proportions depending of uranium enrichment. In fact, the 238 U absorption cross section is very noisy in the range 1-500 eV (see further references) and it is not easy to precisely evaluate the so-called resonance escape probability factor p, i.e. the probability for a neutron not to be captured in this energy range. In the case of a homogeneous core, Uranium fuel and moderator are supposed to be intimately mixed. The outline of this paper is as follows 1. first form of the neutron slowing down equation 2. Second form of the neutron slowing down equation 3. Approximation of the exact transition probability in the case A > 1 − 4. Monte-Carlo analysis of the replacement of the by the − − 5. Exact solution vs Monte-Carlo solution for the − 6. Benefits of an exact solution to the neutron slowing down equation. 7. Reduction of the number of groups with self-shielding. 8. Computation with 652 groups data. 9. Non linear averaging on unit lethargy groups.
2. First form of the neutron slowing down equation. Rather than using the neutron energy as a parameter, we shall use the lethargy defined as = Log (E / E) where E = 10 MeV. 0 0 If Σ ( ) denotes the scattering cross section and Σ ( ) the absorption cross section in the core of a s a nuclear reactor assumed to be homogeneous, then we have Σ( ) = Σ ( ) + Σ ( ), so that in case of s a collision the neutron is absorbed with a probability Σ ( )/ Σ( ) or reappears at another lethargy ’ > a with a probability Σ ( )/ Σ( ). s The probability density for the new lethargy ’ is p( , ’) and is such that , = 1 If the moderator is made with an atom whose atomic mass is equal to A, then it is well known that
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Bertrand Mercier - Arthur Peng