Department of physics

Seminar – 4th grade

Dosimetry

Author: Marko Fajs

Mentor: mag. Denis Glavič-Cindro

Co-mentor: doc. dr. Matej Lipoglavšek

Ljubljana, may 2011

Abstract

This seminar is dealing with fundamentals of radiation dosimetry such as types and sources of ionizing radiation and quantities for describing it. The seminar also shows some actual dose calculations from radionuclides for different pathways. At the end it deals with contributions to world average annual dose from natural background (which is 2,4 mSv).

i

Index 1 Interaction of ionizing radiation with matter .................................................................................. 1

1.1 Alpha radiation ........................................................................................................................ 1

1.2 Beta radiation .......................................................................................................................... 2

1.3 Gamma radiation ..................................................................................................................... 2

2 Quantities describing ionizing radiation with matter...................................................................... 3

2.1 Absorbed dose ......................................................................................................................... 4

2.2 Equivalent dose ....................................................................................................................... 4

2.3 Effective dose .......................................................................................................................... 5

3 Estimation of total individual doses ................................................................................................ 5

3.1 Calculation of external doses from deposited activity ............................................................ 6

3.2 Calculation of external doses from airborne radionuclides .................................................... 7

3.3 Calculation of internal doses due to intake by inhalation or ingestion .................................. 7

3.3.1 Irradiation from inhaled radionuclides ............................................................................ 7

3.3.2 Ingestion of radionuclides ............................................................................................... 7

4 Radiation from natural sources or natural background .................................................................. 7

4.1 Radioactive nuclei in Earth’s crust........................................................................................... 8

4.2 Leakage of radon from the ground ......................................................................................... 9

4.3 Cosmic rays .............................................................................................................................. 9

4.4 Radionuclides in our own body ............................................................................................. 10

4.5 Man-made radionuclides ...................................................................................................... 11

5 Conclusion ..................................................................................................................................... 12

6 References ..................................................................................................................................... 12

ii

Preface Radiation is transfer of energy in the form of a stream of particles or electromagnetic waves. The concept of ionizing radiation covers the flow of elementary particles or nuclei or electromagnetic waves, which have a high enough energy for ionization of atoms or molecules. In the event that a particle or electromagnetic radiation does not have high enough energy for ionization, we are dealing with non-ionizing radiation.

In this seminar I assumed that the reader is familiar with the basics of nuclear physics. Examples of ionizing particles are alpha particles, beta particles and neutrons, while not any electromagnetic radiation is ionizing. Since the energy needed to cause a valence electron to escape an atom is of the order of 4-25 eV, photons must carry kinetic energies in excess of this magnitude to be called “ionizing radiation”. This criterion would seem to include electromagnetic radiation with wavelengths up to about 320 nm, which includes most of the ultraviolet (UV) radiation band (~ 10-400 nm). However, for practical purposes these marginally ionizing UV radiations are not usually considered in the context of radiological physics, since as visible lights they are not capable to penetrate through matter, while other ionizing radiations are generally more penetrating. Electromagnetic spectrum is illustrated on figure 1 [8].

Figure 1: Electromagnetic radiation spectrum (metric scale).

Although ionizing radiation had been present in nature throughout human history (cosmic

radiation, naturally occurring radioactive elements, etc.), we have been actually aware of it for only just over 100 years. Humans do not possess senses that would detect ionizing radiation, contrary to the case of visible light.

As the impact of ionizing radiation on biological system (e.g. human) is important, measurements of ionizing radiation are undertaken, either to monitor or to predict the biological, physical or chemical effects it produces. These effects can only occur with the transfer of energy from the radiation to some irradiated material. The effect is likely to be different if a particular amount of energy is imparted to a small mass of material rather than being distributed throughout a large mass. Most widely used dosimetric quantity for describing these effects is called absorbed dose. Towards the end there are some more words said about radionuclides dosimetry and illustrations on actual assessment of doses for some intake paths for radionuclides. At the end, this seminar concentrates on the dose contributions of naturally occurring ionizing radiation. Artificial (man-made) sources occurring in nature will be mentioned too.

1

1 Interaction of ionizing radiation with matter Ionizing radiation is generally characterized by its ability to excite and ionize atoms of matter with which it interacts.

The range of kinetic or photon energies most frequently encountered extends from some keV to 10 MeV (in practice usually to 3 MeV) and relevant tabulations of data on their interactions with matter tend to emphasize that energy range.

The ICRU (International Commission on Radiation Units and Measurements, 1971) has recommended certain terminology in referring to ionizing radiations which emphasizes the gross differences between the interactions of charged and uncharged radiations with matter:

1. Directly Ionizing Radiation. Moving charged particles have an electrical field surrounding them, which interacts with the atomic structure of the medium i.e. deliver their energy to matter directly, through many small Coulomb-force interactions along the particle’s track. These interactions decelerate the particle and accelerate electrons in the atoms of the medium. The accelerated electrons may acquire enough energy to escape from the parent atom. This process, whereby radiation "strips" off orbital electrons, is called ionization.

2. Indirectly Ionizing Radiation. X-or γ-ray photons or neutrons (i.e., uncharged particles) have no electrical field. They first transfer their energy to charged particles (in the matter, through which they pass) by such means as collisions or scattering. A photon can lose energy by the photoelectric effect, Compton effect, or pair production. The resulting fast charged particles then in turn deliver the energy to the matter as above.

The reason why so much attention is paid to ionizing radiation, and that an extensive science dealing with these radiations and their interactions with matter has evolved, stems from the unique effects that such interactions have upon the irradiated material. Biological systems (e.g., humans) are particularly susceptible to damage by ionizing radiation, so that the expenditure of a relatively trivial amount of energy (~ 4 J/kg) throughout the body is likely to cause death, even though that amount of energy can only raise the gross temperature by about 0,001 °C. Clearly the ability of ionizing radiations to impart their energy to individual atoms, molecules, and biological cells has a profound effect on the outcome. The resulting high local concentrations of absorbed energy can kill a cell either directly or through the formation of highly reactive chemical species such as free radicals in the water medium that constitutes the bulk of the biological material. Ionizing radiations can also produce gross changes, either desirable or deleterious, in organic compounds by breaking molecular bonds, or in crystalline materials by causing defects in the lattice structure.

Even structural steel will be damaged by large enough numbers of fast neutrons, suffering embrittlement and possible fracture under mechanical stress.

1.1 Alpha radiation Alpha radiation is normally produced from the radioactive decay of heavy nuclides and from certain nuclear reactions. The alpha particle consists of 2 neutrons and 2 protons, which equals the nucleus of a helium atom. Because it has no electrons, the alpha particle has a charge of +2 e. This positive charge causes the alpha particle to strip electrons from the orbits of atoms in its vicinity. As the alpha particle passes through material, it removes electrons from the orbits of atoms it passes near. Energy is required to remove electrons and the energy of the alpha particle is reduced by each reaction. Eventually the particle will expend all its kinetic energy, gain 2 electrons in orbit and become a helium atom. Because of its strong positive charge and large mass, the alpha particle deposits a large amount of energy in a short distance of travel. This rapid, large deposition of energy limits the penetration of alpha particles. Alpha particles with the same initial energy make on average the same path through matter. This path length is called the average range for alpha articles. The range

2

depends on particle energy and of the medium itself. Range in the air for alpha particles with energy of 1 MeV is about 1 cm and in water about 0,01 mm.

1.2 Beta radiation A beta-minus particle (β-) is an electron that has been ejected at a high velocity from an unstable nucleus. An electron has a small mass and an electrical charge of -1 e. Beta particles cause ionization by displacing electrons from atom orbits. The ionization occurs from collisions with orbiting electrons. Each collision removes kinetic energy from the beta particle, causing it to slow down. Eventually the beta particle will be slowed enough to allow it to be captured as an orbiting electron in an atom. Although more penetrating than the alpha, the beta is relatively easy to stop and has a low power of penetration. Even the most energetic beta radiation can be stopped by a few millimeters of metal. Figure 2 shows range of β- rays in some materials, which is energy-depended [8].

Positively charged electrons are called positrons (β+). Except for the positive charge, they are identical to beta-minus particles and interact with matter in a similar manner. Positrons are very short-lived, however, and are quickly annihilated by interaction with a negatively charged electron as soon as they are slowed down. The annihilation produces at least two γ-rays (conservation of linear momentum) with a combined energy (1,02 MeV) equal to the rest mass of the positive and negative electrons.

Figure 2: β-rays ranges in some materials.

1.3 Gamma radiation Gamma radiation is electromagnetic radiation. The photon commonly referred to as a γ-ray and is very similar to an x-ray. The difference is that γ-rays are emitted from the nucleus of an atom, and x-rays are produced by de-excitation of electrons in atomic shells. The x-ray is produced when excited electrons move from outer shells to low-energy inner shells. X-rays are also produced when fast-

3

moving electrons approaching an atom are deflected and decelerated as they react with the atom's electrical field (called Bremsstrahlung).

The γ-ray is produced by the decay of excited nuclei and by nuclear reactions. Because the gamma ray has no mass and no charge, it is difficult to stop and has a very high penetrating power. A small fraction of the original gamma stream will still pass through several decimeters of concrete or several meters of water.

There are three main methods of interaction of γ-rays with matter. The first method is referred to as the photo-electric effect (photo effect). When gamma strikes an atom, the total energy of the gamma is expended in ejecting an electron from an inner shell. The result is ionization of the atom and ejection of an electron. This reaction is most dominant with low energy γ-rays interacting in materials with high atomic weight and rarely occurs with gammas having energy above 1 MeV. Any γ-ray energy in excess of the binding energy of the electron is carried off by the electron in the form of kinetic energy.

The second method of attenuation of γ-rays is called Compton scattering. The γ-ray interacts with a bonded or free electron; however, in this case, the photon loses only a fraction of its energy. The actual energy loss depends on the scattering angle. The emitting γ-ray has lower energy, and the energy difference is absorbed by the electron. This reaction becomes important for gamma energies of about 0.1 MeV and higher.

At higher energy levels, a third method of interaction is predominant. This method is pair-production. When a high energy γ-ray passes close enough to a heavy nucleus, it completely disappears, and an electron and a positron are formed. This reaction can take place if the original photon has at least 1,02 MeV energy. Any energy greater than 1,02 MeV becomes kinetic energy shared between the electron and positron. The probability of pair-production increases significantly for higher energy γ-rays.

The weakening of the intensity (absorption) of γ-rays radiation in matter is described by the attenuation coefficient μ, which is energy dependent and is given by

𝐼 𝑥,𝐸 = 𝐼0𝑒−𝜇 (𝐸)𝑥 . (1.1)

Contributions to attenuation coefficient for mentioned interactions with matter for electromagnetic radiation are illustrated in figure 3.

Figure 3: Attenuation coefficient for photons.

2 Quantities describing ionizing radiation with matter Previous chapter shortly described types and interactions of ionizing radiation. In order to monitor them, we need some appropriate quantities for measuring. Such appropriate quantities are:

4

- The absorbed dose, D, describing the energy imparted to matter by all kinds of ionizing radiations, but delivered by the charged particles.

- The equivalent dose, HT, which takes into account different types of radiation. - The effective dose, HE, which takes into account different tissues or organs.

2.1 Absorbed dose From the activity of some radionuclides we cannot make conclusions on health risks for living beings. Health risks depend on the energy of radiation as well as on the type of radiation. Because of this, it is convenient to introduce the quantity absorbed dose.

The absorbed dose is relevant to all types of ionizing radiation fields, whether directly or indirectly ionizing, as well as to any ionizing radiation source distributed within the absorbing medium.

The absorbed dose, D, can best be defined in terms of the related stochastic quantity energy imparted, E. The energy imparted by ionizing radiation to matter of mass m in a finite volume V is defined as

(2) 𝐸 = 𝑅𝑖𝑛 𝑢 − 𝑅𝑜𝑢𝑡 𝑢 + 𝑅𝑖𝑛 𝑐 − 𝑅𝑜𝑢𝑡 𝑐 + Σ𝑄, (2.1)

where (Rin)u is radiant energy of uncharged particles entering V, (Rout)u is the radiant energy of all the uncharged radiation leaving V, (Rin)c is the radiant energy of the charged particles entering V, (Rout)c is the radiant energy of the charged particles leaving V and ΣQ is net energy derived from rest mass in V (if mass goes to energy, then this term is positive but if we gain energy from mass, then the term is negative). We can now define the absorbed dose, D, at any point P in V as

𝐷 =𝑑𝐸

𝑑𝑚, (2.2)

where E is now the expectation value of the energy imparted in the finite volume V during some time interval, dE is that for an infinitesimal volume dV at point P, and dm is the mass of dV. In general we are dealing with non-homogenous matter (tissue). Thus the mean dose, DT, in a specified tissue T is then given by the integral form [10]:

𝐷𝑇 =1

𝑚𝑇 𝐷𝑑𝑚 =

1

𝑚𝑇 𝐷 𝑥,𝑦, 𝑧 𝜌𝑑𝑉, (2.3)

where mT is the mass of the tissue or organ, D is the absorbed dose in the mass element dm (see equation (2.2)), ρ is the density and dV is the volume element.

Thus the absorbed dose, D, is the expectation value of the energy imparted to matter per unit mass at a point. The unit of absorbed dose is Jkg-1 or Gray (Gy). It should be recognized that D represents the energy per unit mass which remains in the matter at P to produce any effects attributable to the radiation. Consequently, the absorbed dose is the most important quantity in radiological physics.

2.2 Equivalent dose The absorbed dose is adequate quantity for measuring ionizing radiation, but it is not enough for describing biological effects, as was told in previous subchapter, health risks are dependent also from types of the radiation. Effects from 1 Gy of alpha particles irradiation are much more deadly than from 1 Gy of γ-rays.

Health risks from different types of ionizing radiation are considered with introduction of radiation quality factor, wR, for the radiation R. The equivalent dose is then defined as

5

𝐻𝑇 = 𝑤𝑅𝐷𝑅

𝑅

. (2.4)

Equivalent dose, HT, has physically the same unit as absorbed dose, but to avoid confusion when multiplying with dimensionless quality factor (wR), new unit is introduced – Sievert (Sv). The radiation quality factors for different types of radiation are given in table 1.

Table 1: Quality factors.

Type and energy range of ionizing radiation wR

Photons, positrons, electrons, muons – all energies 1

Protons (except of thruster protons) – energy above 2 MeV 5

Neutrons – energy under 10 keV 5

Neutrons between 10 and 100 keV 10

Neutrons above 100 keV up to 2 MeV 20

Neutrons above 2 MeV up to 20 MeV 10

Neutrons above 20 MeV 5

Alpha particles, fission products, heavy nuclei 20

2.3 Effective dose Effect, the radiations have on living beings, depends also on tissue through which the radiation passes. Effective dose additionally takes into account the fact that organs or tissues are differently susceptible to ionizing radiations and cancer forming. Effective dose is defined as sum of equivalent doses in particular tissue multiplied by dimensionless tissue weighting factor, wT:

𝐻𝐸 = 𝑤𝑇

𝑇

𝑤𝑅𝐷𝑇,𝑅

𝑅

. (2.5)

Effective dose, HE, also has Sievert for unit. Tissue weighting factors are given in table 2. Table 2: Tissue weighting factors.

Tissue or organ Tissue weighting factor

Gonads 0,20

Red bone marrow 0,12

Lungs 0,12

Stomach 0,12

Colon 0,12

Breasts 0,05

Thyroid gland 0,05

Bladder 0,05

Liver 0,05

Gullet 0,05

Skin 0,01

Periosteum 0,01

Other organs 0,05

Whole body 1,00

3 Estimation of total individual doses The following short chapters will describe or more accurate just illustrate how individual doses are calculated for the transfer of particular radionuclides through some particular pathways for

6

radiological protection (adopted from IAEA, [4]). For generic assessment purposes the total hypothetical critical group dose due to a particular source is estimated by summing the doses from all pathways and all radionuclides. Thus the hypothetical critical group is assumed to represent those members of the public most exposed from the source from all possible pathways (see figure 4). In practice this is unlikely to occur, although it is a reasonable assumption for generic purposes. If more detailed assessments are needed, one uses various models and methodologies for those purposes.

The calculations are based either on the assumption that equilibrium is reached (e.g. constant supply by air or water pathways) or on the assumption of a continuous buildup of long lived radionuclides in the environment (e.g. soil or sediment pathways), as appropriate.

Figure 4: Exposure pathways

3.1 Calculation of external doses from deposited activity The annual effective dose from ground deposition, Hgr (Sv/a, Sievert per year), is given by

(3) 𝐻𝑔𝑟 = 𝐶𝑔𝑟 ,𝑖𝑕𝑔𝑟 ,𝑖𝑂𝑓 ,𝑖

𝑖

, (3.1)

where Of is the fraction of the year for which the hypothetical critical group member is exposed to this particular pathway, hgr is the dose coefficient for exposure to ground deposits (Sv a-1 per Bq m-2, tabulated in [4]), Cgr is the deposition density of a radionuclide i (Bq m-2). If there are more radionuclides that need to be considered, then we make summation of all contributions.

Cgr can be obtained from the ground deposition rate 𝑑 𝑖 calculated according to this equation:

𝐶𝑔𝑟 = 𝑑 𝑖 1− 𝑒(−𝜆𝐸𝑡)

𝜆𝐸𝑖

, (3.2)

where λE is the effective rate constant for reduction of the activity in the top layers of soil (10-20 cm) and t is the duration of the discharge of radioactive material.

7

3.2 Calculation of external doses from airborne radionuclides

The annual effective dose from immersion in the atmospheric discharge cloud, Him (Sv/a) is given by

𝐻𝑖𝑚 = 𝐶𝐴,𝑖𝑕𝑖𝑚 ,𝑖𝑂𝑓 ,𝑖

𝑖

, (3.3)

where CA is the annual average concentration of nuclide i in air (Bq m-3), him is the effective dose coefficient for immersion (Sv a-1 per Bq m-1) and Of is defined the same as above. Dose coefficients are tabulated in [4]. Summation applies in the same case as above.

3.3 Calculation of internal doses due to intake by inhalation or ingestion

3.3.1 Irradiation from inhaled radionuclides The annual effective dose from inhalation, Hinh, is

𝐻𝑖𝑛𝑕 = 𝐶𝐴,𝑖𝑕𝑖𝑛𝑕 ,𝑖𝑅𝑖𝑛𝑕 ,𝑖

𝑖

, (3.4)

where CA is the radionuclide concentration in air (Bq m-3), hinh is the inhalation dose coefficient (Sv Bq-1, tabulated in [4]) and Rinh is the inhalation rate (m3 a-1, also tabulated in [4]). Summation applies as in case above.

3.3.2 Ingestion of radionuclides The ingestion doses are calculated using following general equation:

𝐻𝑖𝑛𝑔 ,𝑝 = 𝐶𝑝 ,𝑖𝑕𝑖𝑛𝑔 ,𝑖𝐹𝑝 ,𝑖

𝑖

, (3.5)

where Hing,p is the annual effective dose from consumption of nuclide i in foodstuff p, Cp,i is the concentration of radionuclide i in foodstuff p at the time of consumption (Bq kg-1), hing is the dose coefficient for ingestion of a radionuclide i (Sv Bq-1) and Fp is the consumption rate for foodstuff p (kg a-1). As before, the default intake rates and dosimetric data are tabulated in [4].

4 Radiation from natural sources or natural background

All living beings including humans are constantly exposed to ionizing radiation from natural sources. Sources contributing to natural background are:

- Radioactive nuclei in Earth’s crust - Leakage of radon from the ground - Cosmic rays - Radionuclides in our own body - Man-made radionuclides1

Based on the UNSCEAR (United Nations Scientific Committee on the Effects of Atomic Radiation) data, the background radiation doses represents the majority of the exposure to ionizing radiation. On average 11 % of the total dose is irradiation from artificial sources among which medical contributions predominate [6].

1 This is quasi natural background, since those contributions results from nuclear testings, accidents

and other human activities.

8

Figure 5: World average contributions to individual effective dose.

4.1 Radioactive nuclei in Earth’s crust

There are radioactive elements in Earth’s crust. Uranium and its daughter products for instance are present in low concentrations in rocks and in soil. Likewise is for thorium and potassium. Radioactive isotopes of these elements decay, emitting radiation comes to surface and consequently we are irradiated mostly by γ-rays. Most houses (i.e. our homes) are built from natural materials, thus we also built some natural radioactivity into our homes. Doses we get from the ground radiation and natural materials in buildings depend on rock and soil types, therefore the values change in different regions. Based on UNSCEAR and NRPB (National Radiological Protection Board) data the average annual dose value we get from the ground is 0,46 mSv. Table 3 shows some average data of radionuclides in construction material (dose contributions from specified values are negligible) for illustration of natural radioactivity [6].

Table 3: Some data on radionuclides in construction material.

Material Uranium Thorium Potassium

ppm Bq/kg ppm Bq/kg ppm Bq/kg

Granite 4,7 63 2 8 4,0 1184

Sandstone 0,45 6 1,7 7 1,4 414

Cement 3,4 46 5,1 21 0,8 237

Limestone cement 2,3 31 2,1 8,5 0,3 89

Sandstone cement 0,8 11 2,1 8,5 1,3 385

Plaster 13,7 186 16,1 66 0,02 5,9

Natural plaster 1,1 15 1,8 7,4 0,5 148

Wood - - - - 11,3 3330

Brick 8,2 111 10,8 44 2,3 666

9

4.2 Leakage of radon from the ground

Radon is a noble gas, which is a daughter product of 238U decay chain. It is spread all over the Earth’s crust. It “finds” the way to the surface and then to the atmosphere, where it continues to decay till it reaches stable form. It is problematic if there is a building on its pathway where it starts to concentrate. Leakage of radon from the ground is illustrated on figure 6. Radon comes to our lungs through inhalation, where it will irradiate us from inside. The world average annual dose is 1,3 mSv but the values can be several times higher in regions with higher radon concentrations.

Figure 6: Radon leakage to a building.

4.3 Cosmic rays

Cosmic rays come from space. Radiation from space can be divided into two groups – primary and secondary. Primary radiation represents particles, especially protons and some other heavier and extremely high energy particles which mostly originate outside our solar system. Only a fraction of these particles penetrates the Earth’s atmosphere. Most of them sooner or later undergo collisions with particles from the atmosphere that produces new particles with lower energies (photons, electrons, neutrons, muons) also called secondary radiation. During these processes some radionuclides which are also found in living tissues are produced and can be detected on surface (3H, 7Be, 14C). Earth’s atmosphere is therefore a good shield against these radiations as it attenuates the cosmic radiation. It is not surprising that with altitude cosmic radiation increases. Earth’s magnetic field also affects the cosmic radiation. Charged particles can penetrate deeper on the poles as they move along the magnetic field if we for instance assume that they are falling perpendicular towards the surface. Contributions from cosmic radiation therefore increase with latitude and altitude. Figure 7 shows average contributions to annual dose from cosmic radiation at different altitudes [6].

10

Figure 7: Average annual doses due to cosmic radiation on different altitudes.

4.4 Radionuclides in our own body

As we need to sustain our life with feeding, we are taking radioactive substances through ingestion into our body. These substances then accumulate in our body and we undergo internal irradiation. The main source of internal irradiation is 40K, which accumulates especially in muscles. Its concentration in the body depends on muscle mass. In a young well-developed sporty man can be twice as large as in old woman. Table 4 shows some radionuclides present in a human body [6].

Table 4: Average values for radionuclides in human body

Radionuclide Radionuclide mass in

the body Total radionuclide

activity Dose

Uranium 90 μg 1,1 Bq 50 nSv

Thorium 30 μg 0,11 Bq 23 nSv 40K 17 mg 4400 Bq 180 µSv

Radium 31 pg 1,1 Bq 0,3 µSv 14C 22 ng 3700 Bq 2 µSv

Tritium 0,06 pg 23 Bq 0,4 nSv

Polonium 0,2 pg 37 Bq 44 µSv

The concentrations of some radionuclides in body presented in above table are natural and do not present significant health risks. It is very hazardous for a human if one ingests or inhalates higher concentrations of radionuclides.

11

4.5 Man-made radionuclides The early atmospheric nuclear weapon’s tests scattered fallout over the whole of the earth's surface and caused the exposure of populations and the contamination of the food chain by a certain number of radionuclides, most of which, given their short radioactive half lives (given in brackets after mentioned isotope), have now vanished. There remain 137Cs (30 years), 90Sr (29 years), some 85Kr (10 years) and 3H (12 years) and the isotopes of plutonium (half lives from 88 years to 24100 years).

In the Chernobyl accident (Ukraine), which occurred in 1986, the total radioactivity dispersed into the atmosphere was of the order of 12∙1018 Bq over a period of 10 days. Three categories of radionuclides were disseminated.

The first consisted of volatile fission products such as 131I (8 days), 133I (20 hours), 134Cs (2 years), 137Cs, 132Te (3 days).

The second was composed of solid fission products and actinides released in much smaller amounts, in particular the strontium isotopes 89Sr (50 days) and 90Sr, the ruthenium isotopes 103Ru (9 days) and 106Ru (368 days) and 239Pu (24100 years). The third category was rare gases which although they represented most of the activity released, were rapidly diluted in the atmosphere. They were mainly 133Xe (5 days) and 85Kr.

The contributions of the early atmospheric nuclear weapon tests and the Chernobyl (category radioactive deposition on the figure 5) accident to the total annual dose are roughly 0.2% (0,005 mSv) and 0.07% (0,002 mSv) respectively. The whole of the nuclear-powered electricity production cycle represents only about 0.007% of total radioactivity. Almost all the radionuclides remain confined inside the nuclear reactors and the fuel cycle plants. 238U, which is non-fissile, can capture neutrons to give in particular plutonium isotopes 239Pu, 240Pu (6560 years) and 241Pu (14 years) and 241Am (433 years). The main fission products generated by the fission of 235U (704 million years) and 239Pu are 131I, 134Cs, 137Cs, 90Sr and 79Se (1 million years).

The main radionuclides present in releases, which are performed in a very strict regulatory framework are, in liquid release, tritium, 58Co (71 days), 60Co, 131I, 134Cs, 137Cs and 110mAg (250 days). In gaseous releases 14C is the most abundant radionuclide, emitted most often as carbon dioxide. In all the reactors in the world, the total production of radiocarbon dioxide amounts to one tenth of the annual production formed naturally by cosmic radiation [9].

Table 5 shows some of the important radionuclides which (may) occur in nature (regardless of the origin) and its dose coefficients [10].

Table 5: Some important radionuclides

Radionuclide Ing. dose coefficient [Sv Bq-1] Inh. dose coefficient [Sv Bq-1] Half life 210Po 1,2 10-6 6,1 10-7 138,38 d 222Rn (noble gas) (noble gas) 3,8 d 137Cs 1,3 10-8 4,6 10-9 30,17 a 90Sr 2,8 10-8 1,6 10-7 28,8 a 210Pb 6,9 10-7 6,1 10-7 22,3 a 131I 2,2 10-8 7,4 10-9 8,02 d 60Co 3,4 10-9 2,1 10-9 5,27 a 58Co 7,4 10-9 3,1 10-8 70,86 d 210Po is around 250,000 times more toxic than hydrogen cyanide (the actual lethal dose for 210Po is about 1 microgram for an 80 kg person compared with about 250 milligrams for hydrogen cyanide). The main hazard is its intense radioactivity (as an alpha emitter), which makes it very difficult to handle safely. Alpha particles emitted by polonium will damage organic tissue easily if polonium is ingested, inhaled, or absorbed (see dose coefficients), although they do not penetrate the epidermis and hence are not hazardous if the polonium is outside the body.

222Rn itself is not dangerous, because it is a noble gas. Radon is hazardous because of its decay products which are not noble gases and can be inhaled. One of its products is the toxic 210Po.

12

137Cesium reacts with water producing a water-soluble compound (cesium hydroxide). After entering the body, cesium gets more or less uniformly distributed throughout the body, with slightly higher concentration in muscle tissues and lower in bones.

90Sr is a "bone seeker" that exhibits biochemical behavior similar to calcium. After entering the organism, most often by ingestion with contaminated food or water, about 70–80 % of the element gets excreted. Virtually all remaining 90Sr is deposited in bones and bone marrow, with the remaining 1 % remaining in blood and soft tissues. Its presence in bones can cause bone cancer, cancer of nearby tissues, and leukemia.

210Pb decays into 210Po. 131I is accumulated in the thyroid, which is very susceptible to ionizing radiation. 60Co decay with beta mode to the stable nickel which is excited and emits two “strong”

gamma rays (1,17 Mev, 1,33 MeV). After entering a human being, most of it gets excreted in feces. A small amount is absorbed by the liver, the kidneys, and the bones, where the prolonged exposure to gamma radiation can cause cancer.

58Co presents health hazards which include cumulative lung damage and dermatitis (beta decay; 1,29 MeV).

5 Conclusion Although ionizing radiation presents health risks, we cannot avoid it. It is present in the nature and all living beings, including humans, adapted to it. One therefore cannot say that it is exposed to the ionizing radiation only because it, for instance, lives near nuclear power plant. Ionizing radiation is part of nature and always will be. But as humans are dealing with ionizing radiations for peaceful intentions, that is for common wealth (electricity), radioactive releases are present and they are under very strict regulatory control to avoid contamination and unwanted additional exposure to ionizing radiation. Actual dose contributions because of these human activities at normal operation are therefore several orders lower than from natural background (see figure 5).

6 References [1] F. H. Attix, Introduction to radiological physics and radiation dosimetry (WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim, 2004). [2] J. R. Greening, Fundamentals of radiation dosimetry (Adam Hilger Ltd, Bristol, 1985). [3] E. B. Podgorsak, Radiation oncology physics: A handbook for teachers and students (International Atomic Enery Agency, Vienna, 2005). [4] Safety Report Series No. 19, Generic Models for Use in Assessing the Impact of Discharges of Radioactive Substances to the Environment (International Atomic Energy Agency, Vienna, 2001). [5] B. Zorko et al., Ovrednotenje rezultatov meritev radioaktivnosti v okolici Nuklearne elektrarne Krško: letno poročilo 2009 (Institut “Jožef Stefan”, Ljubljana, 2010). [6] I. Mele, Raopis: Vse o sevanju, kar ne veste, pa bi želeli (Agencija za radioaktivne odpadke, Ljubljana, junij 2006, št. 14). [7] M. Fajs, Seminarska naloga: Kolikšna je aktivnost posameznih snovi po načelih izključitve, izvzetja in odprave nadzora (Agencija za radioaktivne odpadke, Ljubljana, 2009). [8] M. Fajs, Seminar: Ionizirajoča sevanja (Fakulteta za matematiko in fiziko, Ljubljana, 2010). [9] Natural and artificial radioactivity [online]. [Cited on 24th of may 2011; 23:27]. Available at: http://www.cea.fr/var/cea/storage/static/gb/library/clefs48/pdfgb/encadreagb.pdf [10] Safety Series No. 115, International Basic Safety Standards for protection against Ionizing Radiation and for the Safety of Radiation Sources (International Atomic Energy Agency, Vienna, 1996).