**Shielding of ionizing radiation**means having

**some material between the source of radiation and you**(or some device) that will

**absorb the radiation**.

**Radiation shielding**usually consists of barriers of lead, concrete, or water. Many materials can be used for radiation shielding, but there are many situations in radiation protection. It highly depends on the type of radiation to be shielded, its energy, and many other parameters. For example, even depleted uranium can be used as good protection from gamma radiation, but on the other hand, uranium is inappropriate shielding of neutron radiation.

Radiation shielding means having **some material between the source of radiation and you** (or some device) that will **absorb the radiation**. The amount of shielding required, the type of material of shielding strongly depends on several factors. We are not talking about any optimization.

In fact, in some cases, inappropriate shielding may even worsen the radiation situation instead of protecting people from the ionizing radiation. Basic factors, which have to be considered during the proposal of radiation shielding, are:

- Type of the ionizing radiation to be shielded
- The energy spectrum of the ionizing radiation
- Length of exposure
- Distance from the source of the ionizing radiation
- Requirements on the attenuation of the ionizing radiation – ALARA or ALARP principles
- Design degree of freedom
- Other physical requirements (e.g., transparence in case of leaded glass screens)

See also: Interaction of Radiation with Matter

See also: Shielding of Ionizing Radiation

## Shielding of Radiation in Nuclear Power Plants

Generally, in the nuclear industry, radiation shielding has many purposes. In nuclear power plants, the main purpose is **to reduce the radiation exposure** to persons and staff in the vicinity of radiation sources. In NPPs, the main source of radiation is conclusively the nuclear reactor and its reactor core. Nuclear reactors are, in general, powerful sources of an entire spectrum of types of ionizing radiation. Shielding used for this purpose is called **biological shielding**.

But this is not the only purpose of radiation shielding. Shields are also used in some reactors to reduce the intensity of gamma rays or neutrons incident on the reactor vessel. This radiation shielding protects the reactor vessel and its internals (e.g., the core support barrel) from excessive heating due to gamma-ray absorption fast neutron moderation. Such shields are usually referred to as **thermal shields**.

See also: Neutron Reflector.

A little strange radiation shielding is usually used to protect the material of reactor pressure vessels (especially in PWR power plants). Structural materials of pressure vessels and reactor internals are damaged, especially by **fast neutrons**. Fast neutrons create structural defects, which in result lead to **embrittlement of material of pressure vessel**. To minimize the neutron flux at the vessel wall, also core loading strategy can be modified. In the “out-in” fuel loading strategy, fresh fuel assemblies are placed at the periphery of the core. This configuration causes high neutron fluence at the vessel wall. Therefore, many nuclear power plants have adopted the “in-out” fuel loading strategy (with** low leakage loading patterns – L3P**). In contrast to the “out-in” strategy, low leakage cores have fresh fuel assemblies in the second row, not at the periphery of the core. The periphery contains fuel with higher fuel burnup and lowers relative power, and a very sophisticated radiation shield.

In nuclear power plants, the central problem is to shield against **gamma rays** and **neutrons** because the ranges of charged particles (such as beta particles and alpha particles) in the matter are very short. On the other hand, we must deal with shielding all types of radiation because each nuclear reactor is a significant source of all types of ionizing radiation.

## Shielding of various types of radiation

## Calculation of Shielded Dose Rate in Sieverts from Contaminated Surface

Assume a surface in which **1.0 Ci of 137Cs contaminates.** Assume that this **contaminant** can be approximated by the **point isotropic source,** which contains **1.0 Ci of ^{137}Cs**, with a half-life of

**30.2 years**. Note that the relationship between half-life and the amount of a radionuclide required to give an activity of one curie is shown below. This amount of material can be calculated using λ, which is the decay constant of certain nuclide:

About 94.6 percent decays by beta emission to a metastable nuclear isomer of barium: barium-137m. The main photon peak of Ba-137m is **662 keV**. For this calculation, assume that all decays go through this channel.

**Calculate the primary photon dose rate**, in sieverts per hour (Sv.h^{-1}), at the outer surface of a 5 cm thick lead shield. Then **calculate the** **equivalent and effective dose rates** for two cases.

- Assume that this external radiation field penetrates
**uniformly**through the whole body. That means: Calculate the**effective whole-body dose rate**. - Assume that this external radiation field penetrates
**only lungs,**and the other organs are completely shielded. That means: Calculate the**effective dose rate**.

Note that, primary photon dose rate neglects all secondary particles. Assume that the effective distance of the source from the dose point is **10 cm**. We shall also assume that the dose point is soft tissue and can reasonably be simulated by water, and we use the mass-energy absorption coefficient for water.

See also: Gamma Ray Attenuation

See also: Shielding of Gamma Rays

**Solution:**

The primary photon dose rate is attenuated exponentially, and the dose rate from primary photons, taking account of the shield, is given by:

As can be seen, we do not account for the buildup of secondary radiation. If secondary particles are produced, or the primary radiation changes its energy or direction, the effective attenuation will be much less. This assumption generally underestimates the true dose rate, especially for thick shields and when the dose point is close to the shield surface, but this assumption simplifies all calculations. For this case, the true dose rate (with the buildup of secondary radiation) will be more than two times higher.

To calculate the **absorbed dose rate**, we have to use the formula:

- k = 5.76 x 10
^{-7} - S = 3.7 x 10
^{10}s^{-1} - E = 0.662 MeV
- μ
_{t}/ρ =^{ }0.0326 cm^{2}/g (values are available at NIST) - μ = 1.289 cm
^{-1}(values are available at NIST) - D = 5 cm
- r = 10 cm

**Result:**

The resulting absorbed dose rate in grays per hour is then:

**1) Uniform irradiation**

Since the radiation weighting factor for gamma rays is equal to one, and we have assumed the uniform radiation field (the tissue weighting factor is also equal to unity), we can directly calculate the equivalent dose rate and the effective dose rate (E = H_{T}) from the absorbed dose rate as:

**2) Partial irradiation**

In this case, we assume partial irradiation of lungs only. Thus, we have to use the **tissue weighting factor**, which is equal to **w _{T} = 0.12**. The radiation weighting factor for gamma rays is equal to one. As a result, we can calculate the effective dose rate as:

Note that if one part of the body (e.g., the lungs) receives a radiation dose, it represents a risk for a particularly damaging effect (e.g., lung cancer). If the same dose is given to another organ, it represents a different risk factor.

If we want to account for the buildup of secondary radiation, then we have to include the buildup factor. The **extended formula** for the dose rate is then:

## Buildup Factors for Gamma Rays Shielding

The **buildup factor** is a correction factor that considers the influence of the scattered radiation plus any **secondary particles** in the medium during shielding calculations. If we want to account for the buildup of secondary radiation, then we have to include the **buildup factor**. The buildup factor is then a multiplicative factor that accounts for the response to the un-collided photons to include the contribution of the scattered photons. Thus, the buildup factor can be obtained as a ratio of the total dose to the response for un-collided dose.

The **extended formula** for the dose rate calculation is:

The ANSI/ANS-6.4.3-1991 Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials Standard, contains derived gamma-ray attenuation coefficients and buildup factors for selected engineering materials and elements for use in shielding calculations (ANSI/ANS-6.1.1, 1991).