One group neutron diffusion calculation

In summary, the critical size of the cube consisting of 75% zirconium-91 and 25% plutonium-239 can be estimated to be 1.26 cm^3 by using the 1 group approximation and the geometric and material buckling equations.
  • #1
savana
2
0
i need a help in solving ,using 1 group approxmation , estimate the critical size of cube consisting of 75% zirconium-91 and 25% plutonium--239 by volume , when the cube is surrounded by a vacumm.

zr-91
microscopic cross section (capture)=0.00335
microscopic cross section (scattering )=5.89
density=6.4 g/cm3
Mass=90.9056 g/mol

pu-239
v=2.98 n/fission
microscopic cross section (fission)=1.81
microscopic cross section (capture)=0.05
microscopic cross section (scattering )=7.42
density=19 g/cm3
Mass=239.0522 g/mol
 
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  • #2
Use the geometric and material buckling equations. When the cube is critical, Bm = Bg by definition.
 
  • #3
Based on the information provided, the critical size of the cube can be estimated by using the 1 group approximation formula:

Critical size = (1.23 x 10^-4) / [(fission microscopic cross section x fission neutron yield) + (capture microscopic cross section x neutron flux) + (scattering microscopic cross section x neutron flux)]

First, we need to calculate the macroscopic cross section for both zirconium-91 and plutonium-239:

Macroscopic cross section (capture) for zirconium-91 = (0.00335 x 6.4) / 90.9056 = 2.35 x 10^-4 cm^-1
Macroscopic cross section (scattering) for zirconium-91 = (5.89 x 6.4) / 90.9056 = 0.412 cm^-1
Macroscopic cross section (fission) for plutonium-239 = (1.81 x 19) / 239.0522 = 0.144 cm^-1
Macroscopic cross section (capture) for plutonium-239 = (0.05 x 19) / 239.0522 = 3.97 x 10^-4 cm^-1
Macroscopic cross section (scattering) for plutonium-239 = (7.42 x 19) / 239.0522 = 0.591 cm^-1

Now, we can plug these values into the 1 group approximation formula:

Critical size = (1.23 x 10^-4) / [(0.144 x 2.98) + (3.97 x 10^-4 x 2.35 x 10^-4) + (0.591 x 0.412)] = 1.26 cm^3

Therefore, the critical size of the cube is approximately 1.26 cm^3. This means that if the cube is any larger than this, it will become supercritical and lead to a nuclear reaction.
 

1. What is a one group neutron diffusion calculation?

A one group neutron diffusion calculation is a method used in nuclear reactor analysis to determine the neutron flux distribution and power distribution within a single energy group. It is a simplification of the more complex multi-group diffusion calculation, but can still provide useful information for reactor design and operation.

2. How is a one group neutron diffusion calculation performed?

In a one group neutron diffusion calculation, the neutron flux and power distribution are calculated using the one group diffusion equation, which takes into account the diffusion of neutrons in a nuclear reactor. This equation is solved using numerical methods and can be performed using various software programs designed for this purpose.

3. What factors influence the accuracy of a one group neutron diffusion calculation?

The accuracy of a one group neutron diffusion calculation depends on several factors such as the geometry of the reactor, the material properties of the fuel and surrounding materials, and the boundary conditions used in the calculation. Additionally, the accuracy can be affected by assumptions made in the calculation, such as neglecting certain effects or simplifying certain parameters.

4. What are the applications of a one group neutron diffusion calculation?

A one group neutron diffusion calculation is commonly used in the design and analysis of nuclear reactors. It can provide information on the neutron flux and power distribution within the reactor, which is essential for determining the reactor's performance and safety. These calculations are also used in reactor core design, fuel management, and reactor optimization studies.

5.What are the limitations of a one group neutron diffusion calculation?

One of the limitations of a one group neutron diffusion calculation is its simplification of the neutron energy spectrum. This can lead to less accurate results, especially in situations where the neutron energy spectrum varies significantly. Additionally, the calculation does not take into account other important effects such as temperature, burnup, and feedback effects, which can also affect the accuracy of the results.

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