## Solutions of the Diffusion Equation – Multiplying Systems

**non-multiplying**. In non-multiplying environment neutrons are emitted by a neutron source situated in the center of coordinate system and then they freely diffuse through media. We are now prepared to consider

**neutron diffusion**in

**multiplying system**, which contains fissionable nuclei (i.e.

**Σ**

_{f }**≠ 0**). In this multiplying system we will also study spatial distribution of neutrons, but in contrast to non-multiplying environment these neutrons can trigger

**nuclear fission reaction**. [/su_accordion]

**diffusion equation**

in various geometries that satisfy the **boundary conditions**. In this equation** ν** is number of neutrons emitted in fission and **Σ _{f}** is macroscopic cross-section of fission reaction.

**Ф**denotes a

**reaction rate**. For example a fission of

^{235}U by thermal neutron yields

**2.43 neutrons.**

It must be noted that we will solve the diffusion equation without any external source. This is very important, because such equation is a **linear homogeneous equation** in the flux. Therefore if we find one solution of the equation, then any multiple is also a solution. This means that the** absolute value** of the neutron flux **cannot possibly be deduced** from the diffusion equation. This is totally different from problems with external sources, which determine the absolute value of the neutron flux.

We will start with simple systems (planar) and increase complexity gradually. The most important assumption is that all neutrons are lumped into a **single energy group**, they are emitted and diffuse at** thermal energy** (**0.025 eV**). Solutions of diffusion equations in this case provides an illustrative insights, how can be the neutron flux distributed in a reactor core.

## Solution for the Infinite Slab Reactor

Let assume a **uniform reactor** (multiplying system) in the shape of a slab of physical **width** **a** in the x-direction and infinite in the y- and z-directions. This reactor is situated in the center at x=0. In this geometry the flux does not vary in y and z allowing us to eliminate the y and z derivatives from ∇^{2}. The flux is then a **function of x only**, and therefore the Laplacian and diffusion equation can be written as:

The quantity **B _{g}^{2}** is called the

**geometrical buckling**of the reactor and depends only on the geometry. This term is derived from the notion that the

**neutron flux distribution**is somehow

**‘‘buckled’’**in a homogeneous finite reactor. It can be derived the geometrical buckling is the negative relative curvature of the neutron flux (

**B**). In a small reactor the

_{g}^{2}= ∇^{2}Ф(x) / Ф(x)**neutron flux**have more concave downward or ‘‘buckled’’ curvature (

**higher B**) than in a large one. This is a very important parameter and it will be discussed in following sections.

_{g}^{2}For x > 0, this diffusion equation has two possible solutions **sin(B _{g}x)** and

**cos(B**, which give a general solution:

_{g}x)**Φ(x) = A.sin(B _{g}x) + C.cos(B_{g }x)**

From finite flux condition (**0≤ Φ(x) < ∞**), that required only reasonable values for the flux, it can be derived, that A must be equal to zero. The term **sin(B _{g}x)** goes to negative values as x goes to negative values and therefore it cannot be part of a physically acceptable solution. The physically acceptable solution must then be:

**Φ(x) = C.cos(B _{g }x)**

where **B _{g}** can be determined from

**vacuum boundary condition**.

The vacuum boundary condition requires the relative neutron flux near the boundary to have a **slope** of **-1/d**, i.e., the flux would extrapolate linearly to** 0 at a distance d** beyond the boundary. This **zero flux boundary condition** is more straightforward and is can be written mathematically as:

If d is not negligible, physical dimensions of the reactor are increased by d and extrapolated boundary is formulated with dimension **a _{e}/2 = a/2 + d** and this condition can be written as

**Φ(a/2 + d) = Φ(a**.

_{e}/2) = 0Therefore, the solution must be **Φ(a _{e}/2) = C.cos(B_{g }.a_{e}/2) = 0** and the values of geometrical buckling, B

_{g}, are limited to

**B**, where n is any

_{g}=^{nπ}/_{a_e}**odd integer**. The only one physically acceptable odd integer is

**n=1**, because higher values of n would give cosine functions which would become negative for some values of x. The solution of the diffusion equation is:

It must be added the **constant C cannot be obtained** from this diffusion equation, because this constant gives the absolute value of neutron flux. In fact the** neutron flux** can have** any value** and the **critical reactor** can operate at any power level. It should be noted the **cosine flux shape** is only in theoretical case in a uniform homogeneous reactor at low power levels (at “**zero power criticality**”).

In power reactor core (at full power operation), the neutron flux can reach, for example, about **3.11 x 10**^{13 }**neutrons.cm**^{-2}**.s**^{-1}**, **but this values depends significantly on many parameters (type of fuel, fuel burnup, fuel enrichment, position in fuel pattern, etc.).

The power level does not influence the criticality (k_{eff}) of a power reactor unless thermal reactivity feedbacks act (operation of a power reactor without reactivity feedbacks is between 10E-8% – 1% of rated power).