Direct Coupling of Energy Groups, Multigroup Neutron Diffusion

In summary: This is due to the fact that for neutron moderators with A > 1, the probability of s-wave downscattering is given by P(E_{i}\rightarrow E_{f}) = \frac{1}{(1-\alpha)*E_{i}}, where \alpha = (\frac{A-1}{A+1})^{2}. However, for A = 1 (as in hydrogen), \alpha = 0 and the distribution goes as 1/E. This means that for hydrogen, the probability of skipping energy groups is uniformly distributed as \frac{1}{E_{i}}.
  • #1
NuclearVision
3
0
This one comes from Duderstadt and Hamilton, Problem 7-3.

In multi-group diffusion theory What percentage of neutrons slowing down in hydrogen will tend to skip energy groups if the group structure is chosen such that [itex]\frac{E_{g-1}}{E_{g}}[/itex]=100= 1/[itex]\alpha_{approx}[/itex].

I know that the probability of a neutron scattering to a lower energy in hydrogen is uniformly distributed as: [itex]\frac{1}{E_{i}}[/itex] (because [itex]\alpha[/itex] = 0 in this case).

My approach was to integrate the probability distribution from 0 to the bottom of the (approximated) energy group which should give me the probability that the final energy is less than [itex]\alpha E_{i}[/itex]:

[itex]\int^{E_{i} \alpha_{approx}}_{0} \frac{1}{E_{i}} dE_{f}[/itex] which lead to a value of [itex]\alpha_{approx}=\frac{1}{100}=1[/itex]%.

Am I doing this right?


It is worth noting that for neutron moderators with A> 1 (anything heavier than hydrogen) the s-wave downscattering PDF is given by:
[itex]P(E_{i}\rightarrow E_{f})=\frac{1}{(1-\alpha)*E_{i}}[/itex]

where:
[itex]\alpha=(\frac{A-1}{A+1})^{2}[/itex]

and [itex]\alpha E_{i}[/itex] is the minimum energy a neutron can scatter to from [itex]E_{i}[/itex] in a collision with a nucleus of mass A.

from this it is clear that if A=1 (as in hydrogen) α = 0 and thus the distribution goes as 1/E.


Thanks!
 
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  • #2
Yes, your approach is correct. Since the group structure is such that \frac{E_{g-1}}{E_{g}} = 1/\alpha_{approx}, then the probability that a neutron will skip energy groups is equal to \alpha_{approx} = 1/100 = 1%.
 

1. What is direct coupling of energy groups in multigroup neutron diffusion?

Direct coupling of energy groups in multigroup neutron diffusion is a method used to solve the neutron diffusion equation, which describes the transport of neutrons in a nuclear reactor. In this method, the energy groups are coupled directly to each other, rather than being solved separately. This allows for a more accurate and efficient calculation of the neutron flux in a reactor.

2. How does direct coupling differ from other methods of solving the neutron diffusion equation?

In traditional methods of solving the neutron diffusion equation, the energy groups are solved separately and then coupled together in an iterative process. Direct coupling, on the other hand, solves all the energy groups simultaneously, resulting in a more accurate and efficient solution.

3. What are the advantages of using direct coupling in multigroup neutron diffusion?

There are several advantages to using direct coupling in multigroup neutron diffusion. Firstly, it allows for a more accurate calculation of the neutron flux in a reactor. Additionally, it can reduce the computational time and resources needed for the calculation. Direct coupling also allows for the inclusion of more energy groups, which can provide a more detailed and precise analysis of reactor behavior.

4. Are there any limitations to using direct coupling in multigroup neutron diffusion?

One limitation of direct coupling is that it may not be suitable for all types of reactor designs. It is most commonly used for reactors with a large number of energy groups and a high degree of heterogeneity. Additionally, the accuracy of the results may be affected by the assumptions and simplifications made in the model.

5. How is direct coupling implemented in practice?

Direct coupling is typically implemented using computer codes that solve the multigroup neutron diffusion equation. These codes require input data such as the geometry and materials of the reactor, as well as the neutron cross-sections for the various energy groups. The calculations are then performed iteratively until a converged solution is reached.

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