**the neutron cross-section**is an effective area that quantifies

**the likelihood**of certain interaction between an incident neutron and a target object. It must be noted this likelihood does not depend on real target dimensions, because we are not describing geometrical cross-section. In the vicinity of target nucleus, neutron is subjected to strong nuclear forces of target nucleons. The interaction strongly depends on many variables, such as type of target nucleus and the neutron energy. For example, the likelihood that a thermal neutron will be absorbed by xenon-135 is about a million times higher than it will be scattered.

At this point we have to distinguish between two basic types of nuclear or neutron cross-sections.

**Microscopic Cross-section**. The effective target area in m^{2}presented by a single nucleus to an incident neutron beam is denoted the microscopic cross section,**σ**. The microscopic cross-sections characterize interactions with single isotopes and are a part of data libraries, such as ENDF/B-VII.1.**Macroscopic Cross-section**. The**macroscopic cross-section**represents the**effective target area of all of the nuclei**contained in the volume of the material (such as fuel pellet). The units are given in**cm**. It is the probability of neutron-nucleus interaction per centimeter of neutron travel. These data are commonly used by codes for reactor core analyses and design. These codes are based on^{-1}**pre-computed assembly homogenized macroscopic cross-sections.**

## Barn – Unit of Cross-section

The cross-section is typically denoted **σ** and measured in units of area [m^{2}]. But a square meter (or centimeter) is tremendously large in comparison to the effective area of a nucleus, and it has been suggested that a physicist once referred to the measure of a square meter as being “as big as a barn” when applied to nuclear processes. The name has persisted and microscopic cross sections are expressed in terms of barns. The standard unit for measuring a nuclear cross section is the **barn**, which is equal to **10 ^{−28} m² or 10^{−24} cm²**. It can be seen the concept of a nuclear cross section can be quantified physically in terms of

**“characteristic target area”**where a larger area means a larger probability of interaction.

### Typical Values of Microscopic Cross-sections

**Uranium 235**is a fissile isotope and its fission cross-section for thermal neutrons is about**585 barns**(for 0.0253 eV neutron). For fast neutrons its fission cross-section is**on the order of barns**.**Xenon-135**is a product of U-235 fission and has a**very large****neutron capture cross-section**(about**2.6 x 10**^{6}**barns**).**Boron**is commonly used as a**neutron absorber**due to the high neutron cross-section of isotope. Its^{10}B**(n,alpha) reaction**cross-section for thermal neutrons is about**3840 barns**(for 0.025 eV neutron).**Gadolinium**is commonly used as a**neutron absorber**due to very high neutron absorbtion cross-section of two isotopesand^{155}Gd.^{157}Gd^{155}Gd has 61 000 barns for thermal neutrons (for 0.025 eV neutron) and^{157}Gd has even 254 000 barns.

See also: JANIS (Java-based Nuclear Data Information Software)

## Microscopic Cross-section

The extent to which neutrons interact with nuclei is described in terms of quantities known as **cross-sections**. **Cross-sections** are used to express the **likelihood of particular interaction** between an incident neutron and a target nucleus. It must be noted this likelihood does not depend on real target dimensions. In conjunction with the neutron flux, it enables the calculation of the reaction rate, for example to derive the **thermal power of a nuclear power plant**. The standard unit for measuring the **microscopic cross-section** (σ-sigma) is the **barn**, which is equal to **10 ^{-28} m^{2}**. This unit is very small, therefore barns (abbreviated as “b”) are commonly used.

**The cross-section σ** can be interpreted as the **effective ‘target area’** that a nucleus interacts with an incident neutron. The larger the effective area, the greater the probability for reaction. This cross-section is usually known as **the microscopic cross-section**.

The concept of the microscopic cross-section is therefore introduced to represent the probability of a neutron-nucleus reaction. Suppose that a thin ‘film’ of atoms (one atomic layer thick) with N_{a} atoms/cm^{2} is placed in a monodirectional beam of intensity I_{0}. Then the number of interactions C per cm^{2} per second will be proportional to the intensity I_{0} and the atom density N_{a}. We define the proportionality factor as the microscopic cross-section σ:

**σ _{t} = C/N_{a}.I_{0}**

In order to be able to determine the microscopic cross section, **transmission measurements** are performed on plates of materials. Assume that if a neutron collides with a nucleus it will either be scattered into a different direction or be absorbed (without fission absorption). Assume that there are N (nuclei/cm^{3}) of the material and there will then be N.dx per cm^{2} in the layer dx.

Only the neutrons that have not interacted will remain traveling in the x direction. This causes the intensity of the uncollided beam will be attenuated as it penetrates deeper into the material.

Then, according to the definition of the microscopic cross section, the reaction rate per unit area is Nσ Ι(x)dx. This is equal to the decrease of the beam intensity, so that:

**-dI = N.σ.Ι(x).dx**

and

**Ι(x) = Ι _{0}e^{-N.σ.x}**

It can be seen that whether a neutron will interact with a certain volume of material depends not only on **the microscopic cross-section** of the individual nuclei but also on **the density of nuclei** within that volume. It depends on the **N.σ factor**. This factor is therefore widely defined and it is known **as the macroscopic cross section**.

The difference between the microscopic and macroscopic cross sections is extremely important. The **microscopic cross section** represents the effective target area of a **single nucleus**, while the **macroscopic cross section** represents the effective target area of **all of ****the nuclei** contained in certain volume.

**Microscopic cross-sections** constitute a key parameters of nuclear fuel. In general, neutron cross-sections are essential for the reactor core calculations and are a part of data libraries, such as ENDF/B-VII.1.

The neutron cross-section is variable and depends on:

**Target nucleus**(hydrogen, boron, uranium, etc.). Each isotop has its own set of cross-sections.**Type of the reaction**(capture, fission, etc.). Cross-sections are different for each nuclear reaction.**Neutron energy**(thermal neutron, resonance neutron, fast neutron). For a given target and reaction type, the cross-section is strongly dependent on the neutron energy. In the common case, the cross section is usually much larger at low energies than at high energies. This is why most nuclear reactors use a neutron moderator to reduce the energy of the neutron and thus increase the probability of fission, essential to produce energy and sustain the chain reaction.**Target energy**(temperature of target material – Doppler broadening). This dependency is not so significant, but the target energy strongly influences inherent safety of nuclear reactors due to a Doppler broadening of resonances.

**Microscopic cross-section varies with incident neutron energy**. Some nuclear reactions exhibit **very specific dependency** on incident neutron energy. This dependency will be described on the example of the radiative capture reaction. The likelihood of a neutron radiative capture is represented by the radiative capture cross section as **σ _{γ}**. The following dependency is typical for radiative capture, it definitely does not mean, that it is typical for other types of reactions (see elastic scattering cross-section or (n,alpha) reaction cross-section).

The capture cross-section can be divided into three regions according to the incident neutron energy. These regions will be discussed separately.

**1/v Region****Resonance Region****Fast Neutrons Region**

## Macroscopic Cross-section

The difference between the **microscopic cross-section** and **macroscopic cross-section** is very important and is restated for clarity. The **microscopic cross section** represents the **effective target area of a single target nucleus** for an incident particle. The units are given in **barns or cm ^{2}**.

While the **macroscopic cross-section** represents the **effective target area of all of the nuclei** contained in the volume of the material. The units are given in **cm ^{-1}**.

A macroscopic cross-section is derived from **microscopic cross-section** and the **atomic number density**:

**Σ=σ.N**

Here **σ**, which has units of m^{2}, is the microscopic cross-section. Since the units of N (nuclei density) are nuclei/m^{3}, the macroscopic cross-section Σ have units of m^{-1}, thus in fact is an incorrect name, because it is not a correct unit of cross-sections. In terms of Σ_{t} (the total cross-section), the equation for the intensity of a neutron beam can be written as

**-dI = N.σ.Σ _{t}.dx**

Dividing this expression by I(x) gives

**-dΙ(x)/I(x) = Σ _{t}.dx**

Since dI(x) is the number of neutrons that collide in dx, the quantity –**dΙ(x)/I(x)** represents the probability that a neutron that has survived without colliding until x, will collide in the next layer dx. It follows that the probability P(x) that a neutron will travel a distance x without any interaction in the material, which is characterized by Σ_{t}, is:

**P(x) = e ^{-Σt.x}**

From this equation, we can derive the probability that a neutron will make its **first collision in dx**. It will be the quantity** P(x)dx**. If the probability of the first collision in dx is independent of its past history, the required result will be equal to the probability that a neutron survives up to layer x without any interaction (~Σ_{t}dx) times the probability that the neutron will interact in the additional layer dx (i.e. ~e^{-Σt.x}).

**P(x)dx = Σ _{t}dx . e^{-Σt.x} = Σ_{t} e^{-Σt.x} dx**

## Mean Free Path

From the equation for the probability of the** first collision in dx** we can calculate **the mean free path** that is traveled by a neutron between two collisions. This quantity is usually designated by the symbol **λ** and it is equal to the average value of x, the distance traveled by a neutron without any interaction, over the interaction probability distribution.

whereby one can distinguish** λ _{s}, λ_{a}, λ_{f}**, etc. This quantity is also known as the

**relaxation length**, because it is the distance in which the intensity of the neutrons that have not caused a reaction has decreased with a factor e.

For materials with high absorption cross-section, the mean free path is **very short** and neutron absorption occurs mostly** on the surface** of the material. This surface absorption is called **self-shielding** because the outer layers of atoms shield the inner layers.

## Macroscopic Cross-section of Mixtures and Molecules

Most materials are composed of several chemical elements and compounds. Most of chemical elements contains several** isotopes** of these elements (e.g. gadolinium with its six stable isotopes). For this reason most materials involve many cross-sections. Therefore, to include all the isotopes within a given material, it is necessary to determine the macroscopic cross section for each isotope and then sum all the individual macroscopic cross-sections.

In this section both factors (different** atomic densities** and different **cross-sections**) will be considered in the calculation of the **macroscopic cross-section of mixtures**.

First, consider the Avogadro’s number N_{0} = **6.022 x 10 ^{23}**, is the number of particles (molecules, atoms) that is contained in the amount of substance given by one mole. Thus if M is the

**molecular weight**, the ratio

**N**equals to the number of molecules in 1g of the mixture. The number of molecules per cm

_{0}/M^{3}in the material of density ρ and the macroscopic cross-section for mixtures are given by following equations:

**N _{i} = ρ_{i}.N_{0} / M_{i}**

Note that, in some cases, the cross-section of the molecule** is not equal** to the sum of cross-sections of its** individual nuclei**. For example the cross-section of neutron elastic scattering of water exhibits anomalies for thermal neutrons. It occurs, because the kinetic energy of an incident neutron is of the order or less than **the chemical binding energy** and therefore the scattering of slow neutrons by water (H_{2}O) is greater than by free nuclei (2H + O).

## Doppler Broadening of Resonances

In general, Doppler broadening is the broadening of spectral lines due to the **Doppler effect** caused by a distribution of kinetic energies of molecules or atoms. In reactor physics a particular case of this phenomenon is the **thermal Doppler broadening of the resonance capture cross sections** of the fertile material (e.g. ^{238}U or ^{240}Pu) caused by **thermal motion of target nuclei** in the nuclear fuel.

The Doppler broadening of resonances is **very important phenomenon**, which **improves reactor stability**, because it accounts for the dominant part of the** fuel temperature coefficient** (the change in reactivity per degree change in fuel temperature) in thermal reactors and makes a substantial contribution in fast reactors as well. This coefficient is also called the **prompt temperature coefficient** because it causes an **immediate response** on changes in fuel temperature. The prompt temperature coefficient of most thermal reactors** is negative**.

See also: Doppler Broadening

## Self-Shielding

It was written, in some cases the amount of absorption reactions is dramatically reduced despite the **unchanged microscopic cross-section** of the material. This phenomena is commonly known as **the resonance self-shielding** and also **contributes to to the reactor stability**. There are two types of self-shielding.

**Energy Self-shielding.****Spatial Self-shielding.**

See also: Resonance Self-shieldingAn increase in temperature from T_{1} to T_{2} causes the broadening of spectral lines of resonances. Although the area under the resonance remains the same, the broadening of spectral lines causes an** increase in neutron flux** in the fuel φ_{f}(E), which in turn increases the absorption as the temperature increases.