How to calculate protection against beta radiation. Practical protection against ionizing radiation

Calculation of protection against alpha and beta radiation

time protection method.

distance protection method;

Barrier (material) protection method;

The external exposure dose from gamma radiation sources is proportional to the exposure time. However, for those sources that can be considered point sources in size, the dose is inversely proportional to the square of the distance from it. Therefore, reduction of the personnel exposure dose from these sources can be achieved not only by using the method of protection by a barrier (material), but also by limiting the operating time (protection by time) or by increasing the distance from the radiation source to the worker (protection by distance). These three methods are used in the organization of radiation protection at nuclear power plants.

To calculate the protection against alpha and beta radiation, it is usually sufficient to determine the maximum path length, which depends on their initial energy, as well as on the atomic number, atomic mass and absorbent density.

Protection from alpha radiation at nuclear power plants (for example, when accepting "fresh" fuel) is not difficult due to the short path lengths in the substance. The main danger of alpha-active nuclides is only with internal irradiation of the body.

The maximum path length of beta particles can be determined by the following approximate formulas, see:

for air - R β =450 E β , where E β is the boundary energy of beta particles, MeV;

for light materials (aluminum) - R β = 0.1E β (at E β< 0,5 МэВ)

R β =0.2E β (at E β > 0.5 MeV)

In the practice of work at nuclear power plants, there are sources of gamma radiation of various configurations and sizes. The dose rate from them can be measured with appropriate instruments or calculated mathematically. In the general case, the dose rate from a source is determined by the total or specific activity, the emitted spectrum and geometric conditions - the size of the source and the distance to it.

The simplest type of gamma emitter is the point source. . It is such a gamma emitter for which, without significant loss of calculation accuracy, one can neglect its size and self-absorption of radiation in it. In practice, any equipment that emits gamma radiation at distances more than 10 times its size can be considered a point source.

To calculate protection against photon radiation, it is convenient to use universal tables for calculating the thickness of protection depending on the radiation attenuation ratio K and the energy of gamma rays. Such tables are given in reference books on radiation safety and calculated on the basis of the formula for the attenuation of a wide beam of photons from a point source in matter, taking into account the accumulation factor.

Barrier protection method (narrow and wide beam geometry). In dosimetry, there are concepts of "wide" and "narrow" (collimated) beams of photon radiation. The collimator, like a diaphragm, limits the amount of scattered radiation entering the detector (Fig. 6.1). A narrow beam is used, for example, in some installations for calibrating dosimetric instruments.

Rice. 6.1. Scheme of a narrow photon beam

1 - container; 2 - radiation source; 3 - diaphragm; four - narrow beam of photons

Rice. 6.2. Attenuation of a narrow beam of photons

The weakening of a narrow beam of photon radiation in the protection as a result of its interaction with the substance occurs according to the exponential law:

I \u003d I 0 e - m x (6.1)

where Iо is an arbitrary characteristic (flux density, dose, dose rate, etc.) of the initial narrow photon beam; I - arbitrary characteristic of a narrow beam after passing through the protection of thickness x , cm;

m - linear attenuation coefficient, which determines the proportion of monoenergetic (having the same energy) photons that have experienced interaction in the protection material per unit path, cm -1.

Expression (7.1) is also valid when using the mass attenuation coefficient m m instead of linear. In this case, the thickness of the protection must be expressed in grams per square centimeter (g / cm 2), then the product m m x will remain dimensionless.

In most cases, when calculating the attenuation of photon radiation, a wide beam is used, i.e., a beam of photons, where scattered radiation is present, which cannot be neglected.

The difference between the results of measurements of narrow and wide beams is characterized by the accumulation factor B:

B \u003d Iwide / Inarrow, (6.2)

which depends on the geometry of the source, the energy of the primary photon radiation, the material with which the photon radiation interacts, and its thickness, expressed in dimensionless units mx .

The law of attenuation for a wide beam of photon radiation is expressed by the formula:

I width \u003d I 0 B e - m x \u003d I 0 e - m width x; (6.3),

where m, m br are the linear attenuation coefficient for narrow and wide photon beams, respectively. m and AT for various energies and materials are given in radiation safety handbooks. If the reference books indicate m for a wide beam of photons, then the accumulation factor should not be taken into account.

The following materials are most often used to protect against photon radiation: lead, steel, concrete, lead glass, water, etc.

Barrier protection method (calculation of protection by layers of half attenuation). The radiation attenuation ratio K is the ratio of the measured or calculated effective (equivalent) dose rate P meas without protection, to the allowable level of the average annual effective (equivalent) dose rate P cf at the same point behind a protective screen of thickness x:

P cf = PD A / 1700 h = 20 mSv / 1700 h = 12 μSv / h;

where P cf is the allowable level of the average annual effective (equivalent) dose rate;

PD A - effective (equivalent) dose limit for group A personnel.

1700 hours - the working time fund of group A personnel for the year.

K \u003d P meas / P cf;

where P meas is the measured effective (equivalent) dose rate without protection.

When determining the extremely important thickness of the protective layer of a given material x (cm) from universal tables, one should know the photon energy e (MeV) and the radiation attenuation factor K .

In the absence of universal tables, an operational determination of the approximate thickness of the shielding can be performed using approximate values ​​of the half-length attenuation of photons in the wide beam geometry. The layer of half attenuation Δ 1/2 is such a thickness of protection that attenuates the radiation dose by 2 times. With a known attenuation factor K, it is possible to determine the required number of half-attenuation layers n and, consequently, the thickness of the protection. By definition, K = 2 n In addition to the formula, we present an approximate tabular relationship between the attenuation multiplicity and the number of half-attenuation layers:

With a known number of layers of half attenuation n, the thickness of the protection x = Δ 1/2 n.

For example, the half attenuation layer Δ 1/2 for lead is 1.3 cm, for lead glass - 2.1 cm.

distance protection method. The dose rate of photon radiation from a point source in vacuum varies inversely with the square of the distance. For this reason, if the dose rate Pi is determined at some known distance Ri , then the dose rate Rx at any other distance Rx is calculated by the formula:

P x \u003d P 1 R 1 2 / R 2 x (6.4)

time protection method. The time protection method (limiting the time an employee is exposed to ionizing radiation) is most widely used in the production of radiation-hazardous work in a controlled access zone (CCA). These works are documented by a dosimetric order, which indicates the permitted time for the work to be carried out.

Chapter 7 REGISTRATION METHODS OF IONIZING RADIATION

To the number technical means protection includes the device of various screens made of materials that reflect and absorb radiation. Screens are arranged both stationary and mobile (Fig. 58).

When calculating protective screens, their material and thickness are determined, which depend on the type of radiation, the energy of particles and quanta, and the required multiplicity of its attenuation. The characteristics of protective materials and experience in working with radiation sources make it possible to outline the preferential areas for the use of one or another protective material.

Metal is most often used for the construction of mobile devices, and building materials (concrete, brick, etc.) - for the construction of stationary protective devices.

Transparent materials are most often used for viewing systems and therefore they must have not only good protective, but also high optical properties. The following materials meet such requirements well: lead glass, lime glass, liquid-filled glass (zinc bromide, zinc chloride);

Lead rubber is used as a protective material against gamma rays.

Rice. 58. Mobile screen

The calculation of protective screens is based on the laws of interaction various kinds radiation with matter. Protection against alpha radiation is not a difficult task, since alpha particles of normal energies are absorbed by a layer of living tissue of 60 microns, while the thickness of the epidermis (dead skin) is 70 microns. A layer of air a few centimeters or a sheet of paper is sufficient protection against alpha particles.

When beta radiation passes through a substance, secondary radiation occurs, therefore, it is necessary to use light materials (aluminum, plexiglass, polystyrene) as protective ones, since the energy of bremsstrahlung increases with increasing atomic number of the material.

For protection against beta particles (electrons) high energy lead shields are used, but the inner lining of the shields must be made of a material with a low atomic number in order to reduce the initial energy of the electrons, and hence the radiation energy arising in the lead.

The thickness of the aluminum protective screen (g/cm2) is determined from the expression

d = (0.54Emax - 0.15),

where Emax is the maximum energy of the beta spectrum of a given radioactive isotope, MeV.

When calculating protective devices, first of all, it is necessary to take into account the spectral composition of radiation, its intensity, as well as the distance from the source at which the maintenance personnel is located, and the time spent in the sphere of radiation exposure.

At present, based on the available calculated and experimental data, attenuation ratio tables are known, as well as various kinds of nomograms that allow determining the thickness of protection against gamma radiation of various energies. As an example, in fig. Figure 59 shows a nomogram for calculating the thickness of lead shielding from a point source for a wide beam of Co60 gamma radiation, which ensures a reduction in the radiation dose to the maximum allowable. On the abscissa axis, the thickness of the protection d is plotted, on the ordinate axis, the coefficient K1 is equal to

(24)

where M is the gamma equivalent of the drug, mg*eq. Ra;

t is the time of operation in the sphere of exposure to radiation, h; R is the distance from the source, cm. For example, it is necessary to calculate the protection from the source of Co60, at M = 5000 mEq Ra, if the attendants are at a distance of 200 cm during the working day, i.e. t = 6 hours.

Substituting the values ​​of M, R and t into expression (24), we determine

According to the nomogram (see Fig. 59), we find that for K1 = 2.5-10-1, the thickness of the lead protection is d = 7 cm.

Another type of nomogram is shown in fig. 60. Here, on the y-axis, the multiplicity of attenuation K is plotted, equal to

K=D0/D

Using expression (23), we obtain

where D0 is the dose generated by the radiation source at a given point in the absence of shielding; D is the dose to be created at a given point after the protection device.

Rice. Fig. 59. Nomogram for calculating the thickness of lead shielding from a point source for a wide beam of Co60 gamma radiation

Suppose it is necessary to calculate the thickness of the walls of the room in which the gamma therapeutic unit is located, charged with the Cs137 preparation in 400 g-eq Ra (M = = 400,000 meq Ra). The nearest distance at which the attendants are located in the neighboring room R = 600 cm. According to sanitary standards in the neighboring rooms in which there are people who are not radioactive substances, the radiation dose should not exceed 0.03 rem / week, or for gamma radiation approximately 0.005 rad per working day, i.e. D = 0.005 rad for t = 6 hours of attenuation, we use formula (23). To evaluate the multiplicity

According to fig. 60 we determine that for K = 1.1. 104, the thickness of the concrete protection is approximately 70 cm.

When choosing a protective material, it is necessary to be guided by its structural properties, as well as the requirements for the size and weight of the protection. For protective covers various types(gamma therapeutic, gamma flaw detection), when mass plays a significant role, the most advantageous protective materials are materials that best attenuate gamma radiation. The greater the density and serial number of the substance, the greater the degree of attenuation of gamma radiation.

Therefore, for the above purposes, lead is most often used, and sometimes even uranium. In this case, the thickness of the protection is less than when using another material, and consequently, the mass of the protective casing is less.

Rice. 60. Nomogram for calculating the thickness of protection against gamma radiation by attenuation factor

When creating stationary protection (i.e., protection of premises in which work is carried out with gamma sources), which ensures the stay of people in neighboring rooms, it is most economical and convenient to use concrete. If we are dealing with soft radiation, in which the photoelectric effect plays a significant role, substances with a large serial number, in particular barite, are added to concrete, which makes it possible to reduce the thickness of the protection.

Water is often used as a protective material for storage, i.e. drugs are lowered into a pool of water, the thickness of which provides the necessary reduction in the radiation dose to safe levels. With water protection, it is more convenient to charge and recharge the unit, as well as carry out repair work.

In some cases, the conditions of work with gamma radiation sources may be such that it is impossible to create stationary protection (when recharging installations, removing a radioactive preparation from a container, calibrating an instrument, etc.). Here we mean that the activity of the sources is low. To protect the operating personnel from exposure, it is necessary to use, as they say, "time protection" or "distance protection". This means that all manipulations with open sources of gamma radiation should be done using long grips or holders. In addition, this or that operation must be performed only for the period of time during which the dose received by the worker does not exceed the norm established by sanitary rules. Such work must be carried out under the control of a dosimetrist. At the same time, unauthorized persons should not be in the room, and the area in which the dose exceeds the maximum allowable during operation must be protected.

It is necessary to periodically monitor protection with the help of dosimetric devices, since over time it may partially lose its protective properties due to the appearance of various imperceptible violations of its integrity, for example, cracks in concrete and barite concrete fences, dents and ruptures of lead sheets, etc.

Calculation of protection against neutrons is carried out according to the corresponding formulas or nomograms. In this case, substances with a low atomic number should be taken as protective materials, because with each collision with the nucleus, the neutron loses its most of its energy, the closer the mass of the nucleus is to the mass of the neutron. For protection against neutrons, water and polyethylene are usually used. There are practically no pure neutron fluxes. In all sources, in addition to neutrons, there are powerful gamma-ray fluxes that are formed during fission, as well as during the decay of fission products. Therefore, when designing protection against neutrons, it is always necessary to simultaneously provide for protection against gamma radiation.

Useful information:

where L 0 - distance from a point source of radiation to the top of a conical surface with an angle of 2 q 0 at the top, m;

P- number of protection layers.

where i = 1, ..., 26;

E i -1 ( n ) - upper limit of the energy group, for neutron radiation, MeV;

E i ( n ) - lower limit of the energy group for neutron radiation, MeV;

E 0 = 10.5 MeV.

Ej-1(g) - upper limit of the energy group for gamma radiation, MeV;

Ej(g) - lower limit of the energy group for gamma radiation, MeV;

where D n - dose rate of neutron radiation;

D g - dose rate of gamma radiation.

where qi- in accordance with the application, a column vector whose constituent elementsi-th column of the matrixQ.

where Z ( k ) - search criterion calculated in accordance with the application;

T i ( k ) - quadratic functional calculated in accordance with the application.

If for everyone i = 1, 2, ..., n+ 1 G i ( k ¢ ) greater than zero, then the optimization of the function T is finished and proceed to calculations by item with the value of the counter of completely completed optimization stagesk. If at least one valueG i ( k ¢ ) is less than zero, then proceed to the calculations according to p.

replace X ( k ¢ ) H on the X ( k ¢ ) n+ 5 and repeat the algorithm, starting from p. with a new value of the counterk¢ = k¢ + 1.

k¢ = k¢ + 1.

replace X ( k ) H on the X ( k ) n+ 5 and repeat the execution of the algorithm, starting from n. with a new value of the counterk = k+ 1.

and go over to calculations by n. fork = k+ 1.

ATTACHMENT 1

Constants needed to calculate engineering doses

* Note. Index i with coefficient m denotes the material of the layer in which secondary gamma radiation is formed, index j indicates the layer material for which the calculation is performed.

APPENDIX 2

Ei, MeV

microrem/s

1/cm 2 × s

Energy group number i

Ei, MeV

microrem/s

1/cm 2 × s

Ei, MeV

To g i,

microrem/s

1/cm 2 × s

S g i,

Energy group number i

Ei, MeV

To g i,

microrem/s

1/cm 2 × s

S g i,

where k = 0 , ..., To.

Group current densityJk in i-th group at each pointrkalso represented as the sum of two components

where k = 0 , ..., To.

Group cross section of radiation-material interactionj-th layer;

The second moment of expansion inside the group scattering cross section for the materialj-th layer;

r k , ( j ) - coordinate of the inner surfacej th layer.

where a k i,b k i,g k i- coefficients of equations;

d k i- the right side of the equations.

where A 1=1-D r 1 /3r 1 ; B1 = 1 - D r 1 /3r 0 ;

federal agency of Education

State educational institution

higher professional education

"Ivanovo State Power Engineering University

named after V.I. Lenin

Department of Nuclear Power Plants

RADIATION SAFETY
AND DOSIMETRY OF EXTERNAL GAMMA RADIATION

Guidelines for the implementation of laboratory work No. 1

Ivanovo 2009

Compiled by: A.Yu. TOKOV, V.A. KRYLOV, A.N. FEARS

Editor V.K. SEMENOV

The guidelines are intended for students studying in the specialty "Nuclear Power Plants and Installations", passing a laboratory workshop in the physics of ionizing radiation. The theoretical material given in section 1 supplements and partially duplicates the material read in lectures.

Approved cycle methodological committee IFF

Reviewer:

Department of Nuclear Power Plants, Ivanovo State Power Engineering University named after V. I. Lenin

RADIATION SAFETY AND DOSIMETRY

EXTERNAL GAMMA RADIATION

Guidelines for laboratory work No. 1

on the course "Protection from radiation"

Compiled by: Tokov Alexander Yurievich,

Krylov Vyacheslav Andreevich,

Strakhov Anatoly Nikolaevich

Editor N.S. Rabotaeva

Signed for publication on 7.12.09. Format 60x84 1/16.

The print is flat. Conv. oven l. 1.62. Circulation 100 copies. Order No.

GOUVPO "Ivanovo State Power Engineering University named after V. I. Lenin"

153003, Ivanovo, st. Rabfakovskaya, 34.

Printed in UIUNL ISUE

1. BASICS OF RADIATION SAFETY

1.1. Biological effect of ionizing radiation

Ionizing radiation, acting on a living organism, causes a chain of reversible and irreversible changes in it, the "trigger" of which is ionization and excitation atoms and molecules of matter. Ionization (i.e., the transformation of a neutral atom into a positive ion) occurs if the ionizing particle (α, β - particle, X-ray or γ - photon) transfers energy to the electron shell of the atom, sufficient to detach the orbital electron (i.e. exceeding the binding energy). If the transferred part of the energy is less than the binding energy, then only the excitation of the electron shell of the atom occurs.

AT simple substances ax, the molecules of which are composed of atoms of one element, the ionization process is accompanied by the recombination process. An ionized atom attaches to itself one of the free electrons that are always present in the medium, and again becomes neutral. The excited atom returns to its normal state by the transition of an electron from an upper energy level to a lower one, and a photon of characteristic radiation is emitted. Thus, the ionization and excitation of atoms of simple substances do not lead to any changes in the physicochemical structure of the irradiated medium.

The situation is different when irradiating complex molecules consisting of a large number of different atoms. (protein molecules and other tissue structures). The direct effect of radiation on macromolecules leads to their dissociation, i.e. to break chemical bonds due to ionization and excitation of atoms. The indirect effect of radiation on complex molecules is manifested through the products of radiolysis of water, which makes up the bulk of the body mass (up to 75%). Due to the absorption of energy, the water molecule loses an electron, which quickly transfers its energy to the surrounding water molecules:

H 2 O = > H 2 O + + e – .

As a result, ions, free radicals, radical ions with an unpaired electron (H, OH, hydroperoxide HО 2 ), hydrogen peroxide H 2 O 2 , atomic oxygen are formed:

H 2 O + + H 2 O = > H 3 O + + OH – + H ;

H + O 2 = > BUT 2 ; BUT 2 + NO 2 => H 2 O 2 + 2O.

Free radicals containing unpaired electrons are extremely reactive. The lifetime of a free radical does not exceed 10 -5 s. During this time, the products of water radiolysis either recombine with each other or enter into catalytic chain reactions with protein molecules, enzymes, DNA, and other cellular structures. Induced by free radicals chemical reactions develop with a high yield and involve in this process many hundreds and thousands of molecules not affected by radiation.

The action of ionizing radiation on biological objects can be divided into three stages occurring at different levels:

1) at the atomic level - ionization and excitation of atoms, occurring over a time of the order of 10 -16 - 10 -14 s;

2) at the molecular level – physical and chemical changes in macromolecules caused by direct and radiolytic action of radiation, leading to disturbances of intracellular structures, for a time of the order of 10 -10 - 10 -6 s;

3) at the biological level - violations of the functions of tissues and organs that develop over a period of several seconds to several days or weeks (with acute lesions) or over years or decades (long-term effects of exposure).

The main cell of a living organism is a cell, the nucleus of which in humans contains 23 pairs of chromosomes (DNA molecules) that carry encoded genetic information that ensures cell reproduction and intracellular protein synthesis. Separate sections of DNA (genes) responsible for the formation of any elementary trait of an organism are located on the chromosome in a strictly defined order. The cell itself and its relationship with the extracellular environment is maintained by a complex system of semi-permeable membranes. These membranes regulate the flow of water, nutrients and electrolytes into and out of the cell. Any damage can threaten the viability of the cell or its ability to reproduce.

Among the various forms of disorders, DNA damage is the most important. However, the cell has a complex system of repair processes, especially within the DNA. If the recovery is not complete, then a viable but altered cell (mutant) may appear. The appearance and reproduction of altered cells can be affected, in addition to irradiation, by other factors that arise both before and after exposure to radiation.

In higher organisms, cells are organized into tissues and organs that perform a variety of functions, for example: the production and storage of energy, muscle activity for movement, the digestion of food and the excretion of waste, the supply of oxygen, the search for and destruction of mutant cells, etc. Coordination of these types of body activities is carried out nervous, endocrine, hematopoietic, immune and other systems, which in turn also consist of specific cells, organs and tissues.

random distribution The acts of absorption of energy created by radiation can damage vital parts of the double helix of DNA and other macromolecules of the cell in various ways. If a significant number of cells in an organ or tissue have died or are unable to reproduce or function normally, the function of the organ may be lost. In an irradiated organ or tissue, metabolic processes are disturbed, the activity of enzyme systems is suppressed, tissue growth slows down and stops, new chemical compounds appear that are not characteristic of the body - toxins. The final unwanted radiation effects are divided into somatic and genetic.

Somatic effects manifest themselves directly in the exposed person or as early detectable effects exposure (acute or chronic) radiation sickness and local radiation injuries), or both long-term effects(reduction in life expectancy, the occurrence of tumors or other diseases), manifesting itself several months or decades after irradiation . Genetic, or hereditary, effects- these are the consequences of irradiation of the genome of germ cells, which are inherited and cause congenital deformities and other disorders in offspring. These effects of exposure can be very long-term and extend over several generations of people.

The intensity of the effect harmful effects depends on the specific irradiated tissue, as well as the body's ability to compensate or repair damage.

The ability to regenerate cells depends from the age of the person at the time of irradiation, on gender, state of health and genetic predisposition of the organism, as well as on the magnitude absorbed dose(radiation energy absorbed per unit mass of biological tissue) and, finally, from type of primary radiation that affects the body.

1.2. Threshold and non-threshold effects in human exposure

In accordance with the modern concepts set forth in ICRP Publication 60 and underlying the Russian Radiation Safety Standards NRB-99, possible harmful effects of exposure to health are divided into two types: threshold (deterministic) and non-threshold (stochastic) effects.

1.Deterministic (threshold) effects - direct early, clinically detected radiation diseases with dose thresholds below which they do not occur, and above - the severity of the effects depends on the dose. These include acute or chronic radiation sickness, radiation cataract, impaired reproductive function, cosmetic damage to the skin, dystrophic damage to various tissues, etc.

Acute radiation sickness occurs after exceeding a certain threshold dose of a single exposure and is characterized by symptoms that depend on the level of the dose received (Table 1.1). Chronic radiation sickness develops with systematically repeated exposure, if single doses are lower than those that cause acute radiation injuries, but significantly higher than the permissible limits. Signs of chronic radiation sickness are changes in the composition of the blood (decrease in the number of leukocytes, anemia) and a number of symptoms from nervous systems s. Similar symptoms occur in other diseases associated with weakened immunity, so it is very difficult to identify chronic radiation sickness if the fact of exposure has not been established for certain.

In many organs and tissues, there is a continuous process of cell loss and replacement. The increase in losses can be compensated by an increase in the rate of replacement, but there may also be a temporary, and sometimes permanent decrease in the number of cells capable of maintaining the function of an organ or tissue.

The resulting cell loss can cause a severe disorder that can be detected clinically. Therefore, the severity of the observed effect depends on the radiation dose and there is a threshold below which cell loss is too small to appreciably impair tissue or organ function. In addition to cell death, radiation can cause tissue damage in other ways: by affecting numerous tissue functions, including the regulation cellular processes, inflammatory reactions, suppression of the immune system, hematopoietic system (red bone marrow). All of these mechanisms ultimately determine the severity of deterministic effects.

The value of the threshold dose is determined by the radiosensitivity of the cells of the affected organ or tissue and the body's ability to compensate or restore such damage. As a rule, the deterministic effects of radiation are specific and do not arise under the influence of other physical factors, and the relationship between the effect and exposure is unambiguous (deterministic). Threshold doses for the occurrence of deterministic effects leading to the imminent death of adults are given in Table 1.2. In the case of long-term chronic exposure, the same effects occur at higher total doses than in the case of a single exposure.

The average dose thresholds for the occurrence of deterministic effects are given in Table. 1.1 - 1.3. The severity of the effect (the degree of its severity)

increases in persons with higher radiosensitivity (children, persons with poor health, persons with medical contraindications to work with radiation sources). For such individuals, the values ​​of the dose thresholds of exposure indicated in Table 1.1 may be 10 or more times lower.

Table 1.1. The impact of various doses of radiation on the health of an adult

with a single irradiation

Note. At present, there are a number of anti-radiation agents and successful experience has been accumulated in the treatment of radiation sickness, which makes it possible to prevent death at doses up to 10 Sv (1000 rem).

Table 1.2. Range of acute exposure leading to human death

The dependence of survival on the radiation dose is characterized by the average absorbed dose D 50/60, at which half of the people will die after 60 days. For a healthy adult, such a dose (averaged over the whole body) is 3 - 5 Gy (Gy) for acute exposure (Table 1.2).

Under production conditions, the occurrence of deterministic effects is possible only in a radiation accident, when the radiation source is in an uncontrolled state. In this case, the exposure of people is limited by taking urgent measures - interventions. The dose criteria adopted in NRB-99 for urgent intervention in the event of a radiation accident are based on data on threshold doses for the occurrence of life-threatening deterministic effects (Table 1.3).

Table 1.3. Threshold doses for the occurrence of deterministic effects

and criteria for urgent intervention in a radiation accident

The established limits of occupational exposure doses are tens and hundreds of times lower than the threshold doses for the occurrence of deterministic effects, therefore the main task of modern radiation safety is to limit the possibility of stochastic effects in humans due to exposure under normal conditions.

2. Stochastic or non-threshold effects - long-term effects of exposure that do not have a dose threshold, the probability of which is directly proportional to the radiation dose, and the severity does not depend on the dose. These include cancers and hereditary diseases that occur spontaneously over the years in people due to a variety of natural causes.

The reliability of the connection of a certain part of these effects with exposure was proved by international medical and epidemiological statistics only in the early 1990s. Stochastic effects are usually detected through long time after irradiation and only during long-term observation of large population groups of tens and hundreds of thousands of people. The average latent period is about 8 years for leukemia and 2-3 times longer for other types of cancer. The risk of dying from cancer due to exposure is not the same for men and women and varies depending on the time after exposure (Fig. 1.1).

The probability of malignant transformation of a cell is affected by the magnitude of the radiation dose, while the severity of a certain type of cancer depends only on its type and localization. It should be noted that if the irradiated cell did not die, then it has a certain ability to self-repair the damaged DNA code. If this did not happen, then in a healthy body its vital activity is blocked by the immune system: the regenerated cell is either destroyed or does not multiply until its natural death. Thus, the probability of oncological disease is small and depends on the "health" of the immune and nervous systems of the body.

The process of reproduction of cancer cells is random, although due to genetic and physiological characteristics, people can vary greatly in sensitivity to radiation-induced cancer. Some people with rare genetic diseases can be significantly more sensitive than the average person.

With small dose additions to natural (background) exposure, the probability of causing additional cancer cases is naturally small, and the expected number of cases that can be attributed to an additional dose in an exposed group of people may be less than 1 even in a very large group of people. Since the natural radiation background always exists, as well as the spontaneous level of stochastic effects, any practical activity that leads to additional exposure also leads to an increase in the probability of stochastic effects. The probability of their occurrence is assumed to be directly proportional to the dose, and the severity of the manifestation is not dependent on the radiation dose.

Figure 1.2 illustrates the relationship between exposure and the incidence of cancer at the population. It is characterized by a significant level of spontaneous cancers in the population and a relatively low probability of occurrence of additional diseases under the influence of radiation. In addition, according to UNSCEAR, the spontaneous incidence and mortality from cancer varies significantly both from country to country and from year to year in one particular country. This means that by analyzing the effects of exposure to radiation on a large group of people exposed to the same dose, it is possible to establish a probabilistic relationship between the radiation dose and the number of additional cancers resulting from exposure, however, it is not possible to determine which disease is a consequence of exposure and which has arisen spontaneously.

Figure 1.3 provides an estimate of the size of a group of equally exposed adults needed to reliably confirm the relationship between the increase in the total number of cancers in the group and the radiation dose. Line A-B in the figure defines the theoretical estimate of the group size required to detect additional stochastic effects of radiation with a confidence interval of 90%. Above this line is an area in which it is theoretically possible to prove a connection between an increase in the number of stochastic effects in a group and exposure. Below this line, it is theoretically impossible to prove this connection. The dotted line shows that in order to reliably detect additional effects from uniform irradiation of the body of adults with photons with a dose of 20 mGy, equal to the occupational exposure dose limit, it is necessary to examine at least 1 million people with such a dose.

Thus, the task of ensuring radiation safety is reduced to: 1) prevention of deterministic effects in workers by controlling radiation sources; 2) to reduce the additional risk of stochastic effects by limiting exposure doses and the number of exposed persons.

1.3. Basic dosimetric quantities and units of their measurement

Activity (A) – a measure of the amount of a radionuclide in a source or in any substance, including the human body. Activity is equal to the rate of radioactive decay of the nuclei of atoms of the radionuclide. The value of the total activity characterizes the potential radiation hazard of the premises in which work with radioactive substances is carried out.

SI unit - Bq(becquerel) equal to 1 disintegration per second ( s -1).

Off-system unit - Key(curie); 1 Ci \u003d 37 GBq \u003d 3.7 × 10 10 s -1.

Particle flow ( F) - number elementary particles(alpha, beta, photons, neutrons) emitted by the source or affecting the target per unit time. Unit of measurement - part / s, photon / s or simply s - 1 .

The type and number of particles (photons) emitted during nuclear transformations are determined by the type of decay of the radionuclide nuclei. Since the direction of particle emission is random, the flow propagates in all directions from the source. The total radiation flux of a source is related to its activity by the relation

where v, % is the particle yield per 100 decays (given in reference books on radionuclides; for different radionuclides, the yield varies significantly, v= 0.01% - 200% or more).

Particle Fluence (F) is the ratio of the number of elementary particles (alpha, beta, photons, neutrons) penetrating into the elementary sphere to the area of ​​the central section of this sphere. Fluence, like dose, is an additive and non-decreasing quantity - its value always accumulates over time. Unit of measurement - part / cm 2, photon / cm 2 or simply cm –2 .

Particle flux density ( j) - fluence per unit of time. Unit of flux density of particles or quanta - cm–2 s–1. The flux density characterizes the level (intensity) of radiation at a given point in space (or the radiation situation at a given point in the room).

Energy (E R ) - is the most important characteristic ionizing radiation. In nuclear physics, an off-system unit of energy is used - the electron volt (eV). 1 eV = 1.6020×10 -19 J.

Exposure dose (X) - a measure of the amount of ionization destruction of atoms and molecules of the body during irradiation. It is equal to the ratio of the total charge of all ions of the same sign, created by photon radiation in air, to the mass of the irradiated air volume. The exposure dose is used only for photon radiation with energies up to 3 MeV. In the field of radiation safety, it has been decommissioned since 1996.

SI unit - C/kg(coulomb per kilogram).

Off-system unit - R(X-ray); 1 P = 2.58×10 -4 C/g; 1 C/kg = 3872 R.

Absorbed dose, or simply dose ( D) - a measure of the physical impact of ionizing radiation on a substance (at the molecular level). It is equal to the ratio of the radiation energy absorbed in the substance for the formation of ions to the mass of the irradiated substance.

SI unit - Gr(grey); 1 Gy = 1 J/kg.

Off-system unit - glad(rad – radiation absorbed dose);

1 rad = 0.01 Gy = 10 mGy.

The exposure dose of photon radiation X = 1Р corresponds to the absorbed dose in air D = 0.87 rad (8.7 mGy), and in biological tissue D = 0.96 rad (9.6 mGy) due to different work ionization of molecules. For practical purposes of radiation safety, it can be considered that 1 R corresponds to 1 rad or 10 mGy.

Equivalent dose (N) - a measure of the biological effect of radiation on an organ or tissue (at the level of living cells, organs and tissues). It is equal to the product of the absorbed dose by radiation weighting factor W R , which takes into account the quality of radiation (linear ionizing power). For mixed radiation, the equivalent dose is defined as the sum of the types of radiation « R » :

H = å D R × W R

Radiation weighting coefficient values W R adopted in NRB-99. For alpha, beta, photon and neutron radiation they are equal:

W a = 20; W b= W g = 1; W n = 5 - 20(W n depends on the neutron energy).

SI unit - Sv(sievert); for gamma radiation 1 Sv = 1 Gy.

Off-system unit - rem(biological equivalent of rad);

1 rem = 0.01 Sv = 10 mSv.

Relationship with other dosage units:

For X-ray, beta and gamma radiation 1 Sv = 1 Gy = 100 rem » 100 R;

For alpha radiation (W R \u003d 20) 1 Gy \u003d 20 Sv or 100 rad \u003d 2000 rem;

For neutron radiation, an absorbed dose of 1 rad (10 mGy) would correspond to an equivalent dose of 5–20 rem (50–200 mSv), depending on the energy of the neutrons.

Effective dose (E) - a measure of the risk of the occurrence of remote stochastic effects (at low doses of radiation), taking into account the unequal radiosensitivity of organs and tissues. With uniform irradiation of the whole body, the effective dose coincides with the equivalent: E = H, where H- the same equivalent dose to all organs and tissues .

In the case of uneven exposure, the effective dose is determined as the sum of the organs and tissues "T" :

E = å H T × W T(T = 1 ... 13),

where H T is the equivalent dose to the organ or tissue "T »; W T – weighting coefficient of radiosensitivity of an organ (tissue) . The values ​​of W T are accepted in NRB-99 for 13 organs (tissues), in total they amount to one (see Table 2.1). Effective dose unit– mSv(millisievert).

Collective dose ( S) is a measure of potential damage to society from the possible loss of man-years of full-fledged life of the population due to the realization of long-term consequences of exposure. Equal to the sum of annual individual effective doses E i received by a team of N people:

S= å E i (i = 1…N).

Unit of measurement - man-Sv(man-sievert).

To justify the costs of radiation protection in NRB-99, it is assumed that exposure to a collective dose of S = 1 man-Sv leads to potential damage equal to the loss of 1 man-year of the working life of the population.

Dose rate ( , , or ) is the time derivative of the corresponding dose value (i.e., the rate of dose accumulation). Directly proportional to the particle flux density j , acting on the body. As well as the flux density, the dose rate characterizes the radiation situation (radiation level) at the point of the room or on the territory.

The following abbreviations are often used:

MD (MPD)– dose rate (absorbed dose) ( 1 µGy/h = 100 µrad/h);

MED is the equivalent dose rate ( 1 µSv/h = 100 µrem/h).

natural background - this is the level of natural gamma radiation, which on average at sea level is due to 1/3 of cosmic rays and 2/3 of the radiation of natural radionuclides contained in the earth's crust and materials. Natural background radiation can be measured in units of photon flux density (j) or in units of dose rate.

The level of natural (background) gamma radiation in open areas in units of exposure dose rate is within = (8–12) µR/h. This corresponds to the flux density j about 10 photons / (cm 2 s), as well as:

In MPD units =(8–12) mcrad/h =(0.08–0.12) µGy/h=(80–120) nGy/h,

In DER units = =(0.08–0.12) µSv/h =(80–120) nSv/h.

In some buildings, due to the increased concentration of natural radionuclides in building materials it is allowed to exceed the DER of natural gamma radiation above the background level in open areas by up to 0.2 µSv/h, i.e. up to (0.25–0.35) µSv/h.

In some places on the globe, the natural background can reach
(0.5–0.6) µSv/h, which should be considered normal.

The annual dose of natural radiation (received in 8760 hours) can thus range from 0.8–1 mSv to 2–6 mSv for different inhabitants of the Earth.

1.4. Basic provisions of the NRB-99 Radiation Safety Standards

Radiation safety standards NRB-99 are used to ensure human safety in all conditions of exposure to ionizing radiation of artificial or natural origin.

According to the possibilities of source control and exposure control, the Norms differ four types of exposure to radiation per person :

· from technogenic sources in the conditions of their normal operation (the source and radiation protection are under control and managed);

the same, in the conditions of a radiation accident (uncontrolled exposure);

from natural sources of radiation (uncontrolled exposure);

from medical sources for the purpose of diagnosing and treating diseases.

The requirements for limiting radiation exposure are formulated in NRB-99 separately for each type of exposure. The total dose from all four types of exposure is not considered.

technogenic called artificial sources specially made by man for useful application radiation(instruments, devices, installations, including specially concentrated natural radionuclides), or sources that are by-products of human activity (for example, radioactive waste).

The requirements of the Rules apply to sources from which exposure can be controlled. From control sources of radiation are released that are not capable of creating an individual annual effective dose of more than 10 μSv and a collective dose of more than 1 man-Sv per year under any conditions of their handling (the risk of increasing stochastic effects at such doses is trivial and does not exceed 10 - 6 1/man-year).

The main goal of radiation safety is the protection of public health, including personnel, from the harmful effects of radiation, without unreasonable restrictions useful activity when using radiation in various areas of the economy, in science and medicine.

To ensure radiation safety during normal operation of sources, three basic principles of RB:

· justification principle – prohibition of all types of activity on the use of radiation sources, in which the benefit received for the person and society does not exceed the risk possible harm caused by additional exposure;

· rationing principle – non-exceeding of admissible limits individual exposure doses of citizens from all sources of exposure;

· optimization principle – maintenance at the lowest possible and achievable level taking into account economic and social factors individual exposure doses and number of exposed persons(in international practice this principle is known as ALARA - As Low As Reasonably Achievable - As low as reasonably achievable).

NRB-99 requirements for limiting man-made exposure under controlled conditions (during normal operation of radiation sources).

1. The following categories of exposed persons are established:

· Group A personnel(persons directly working with technogenic sources);

· Group B staff(persons who, according to the working conditions, are in the sphere of their influence);

· population (all persons, including personnel outside the scope and conditions of production activities).

Group A personnel include persons at least 20 years of age who do not have medical contraindications for working with ionizing radiation, who have undergone special training and subsequently undergo an annual medical examination. Group B personnel - persons under 18 years of age (including students undergoing laboratory practice with sources). In the “Population” category, as a rule, children aged 0 years and over are singled out. Many concepts in NRB-99 are standardized, for example, the average life expectancy when considering the risk of non-threshold effects is taken equal to 70 years.

· basic dose limits (PD) – such values ​​of the individual annual effective dose, the non-exceedance of which guarantees the complete exclusion of threshold deterministic effects, and the probability of stochastic non-threshold effects does not exceed the risk acceptable to society;

· permissible levels (DU) are derivatives of the main dose limits for assessing the radiation situation. At one-factor exposure from external sources is the average annual allowable dose rate in working premises ( DMD );

· reference levels (CL) – the levels of exposure doses, activities, flux densities, etc., actually achieved in the organization, ensuring the reduction of personnel exposure as low as reasonably achievable through radiation protection measures.

3. Basic dose limits (PD) do not include doses from natural and medical exposure, as well as doses due to radiation accidents. These types of exposure are subject to special restrictions. The AP values ​​for the categories of exposed persons are given in Table 1.4, and Table 1.5 shows the AMD values ​​for the standard annual exposure time.

4. Effective dose of personnel exposure for 50 years of the period labor activity should not exceed 1000 mSv, and for the population over a period of life of 70 years - 70 mSv.

5. With simultaneous exposure of a person to sources of external and internal radiation (multifactorial irradiation) the main dose limits indicated in Table 1.4 refer to total annual dose due to all factors. Therefore, the values ​​of DU (DMA) for each exposure factor separately should be taken less than in Table 1.5.

6. For women under the age of 45, assigned to group A personnel, additional restrictions have been introduced: the equivalent dose to the lower part of the abdomen should not exceed 1 mSv per month. Under these conditions, the effective dose of irradiation of the fetus for 2 months. undetected pregnancy will not exceed 1 mSv. After establishing the fact of pregnancy, the administration of the enterprise is obliged to transfer the woman to a job not related to radiation.

7. Planned increased exposure above the established dose limits (PD = 50 mSv in terms of effective dose) is allowed during the liquidation or prevention of an accident only if it is necessary to save people and (or) prevent their exposure. Such irradiation is allowed only for men over 30 years of age only with their voluntary written consent, after being informed about the possible doses and health risks. Irradiation in doses up to 2 PD (100 mSv) or up to 4 PD (200 mSv) is allowed only with the permission of the territorial or federal bodies of the State Sanitary and Epidemiological Supervision, respectively, and only for persons classified as group A personnel.

8. Exposure in doses above 4 PD (200 mSv) regarded as potentially dangerous. Persons exposed to radiation in such doses, subsequent work with radiation sources is allowed only on an individual basis by decision of the competent medical commission.

cases unplanned increased exposure in humans at doses above the exposure limit are subject to investigation.

Table 1.4. Basic dose limits

**All values ​​of PD and DU for group B personnel are equal 1 / 4 from the corresponding values ​​for group A personnel.

Table 1.5. Permissible levels for single-factor external exposure

2.1. Preparation for work

Objective

1. Assessment of radiation safety of students and laboratory staff when working with a sealed radionuclide source of gamma radiation.

2. Study of the law of attenuation of gamma radiation with distance from the source.

3. Verification of the readings of various dosimeters with the calculation of the dose rate.

Applied equipment and materials

1. Closed radionuclide source of gamma radiation with the isotope 27 Co 60 (cobalt-60), placed in a protective container made of lead with a wall thickness of 10 cm. collimator(an opening channel that makes it possible to obtain a limited beam of g-radiation).

2. A mobile carriage and a ruler with divisions for measuring the distance from the source to the measuring sensor (detector).

3. Dosimeters with detectors that register gamma radiation.

The main characteristics of the installation with a source of gamma radiation

Term "sealed radionuclide source" means technical product, the design of which excludes the spread of radioactive substances to the environment under the conditions of use and wear for which it is designed. Gamma source cobalt GIK-2-9 is a sealed stainless steel capsule (cylinder 10 x 10 mm), inside which is a radioactive isotope Co-60. A useful flow of gamma quanta freely penetrates through the thin walls of the capsule (with little filtration). For the purposes of this work, the source can be considered as a point, isotropic, and monoenergetic source.

To protect against gamma radiation, the GIK-2-9 source is placed in a lead container with a wall thickness x = 10.5 cm, which has a through collimating channel closed with a lead plug. When the plug is removed, a slightly expanding working beam of gamma radiation is obtained, directed away from people. In this beam, measurements of the dose rate are made at various distances from the source.

In the report on working from the laboratory poster, you must write out:

sketch of a protective container with a source (in section);

photon energy of cobalt gamma radiation (Еg = 1.25 MeV);

The half-life of the Co-60 isotope (T 1/2 = 5.27 years);

the initial activity of the source ao(Bq) and date of attestation of the source;

Passport exposure dose rate at a distance of 1 m (µR/h);

the value of the gamma constant of cobalt-60 G (nGy × m 2 / (s × GBq))

2.2. Assessment of radiation safety when working with a source

Persons staying in the dosimetry laboratory, by order of the university, are classified as "group A personnel" (teachers and employees) and "group B personnel" (students). Permissible limits of the annual effective dose according to NRB-99 for them are, respectively, PD A = 20 mSv and PD B = 5 mSv.

To assess radiation safety, one should estimate the annual effective dose of a worker, separating the man-made component from the natural one. For such measurements, the most suitable portable digital dosimeter MKS-08, included in the mode of measuring the equivalent dose rate (µSv/h). Attention: to obtain correct readings, the device should be directed with the detector (back side of the housing) towards the radiation source.

1. Having walked around the laboratory room with a dosimeter, perform radiation reconnaissance, i.e. find places with high levels of gamma radiation. It is recommended to measure DER on the surface of all devices marked with radiation hazard signs(containers, safes, sets of sources on other desktops). Record DER values ​​for 3-4 characteristic points in the report, indicating them on the floor plan.

2. Determine the average value of the natural background (equivalent dose rate f) at points located at the maximum distance from man-made sources, and also, if possible, outside the window (in this case, pay attention to the difference in readings outside the window and inside the room).

3. Measure the average value of the equivalent dose rate rm at the workplace, located as close as possible to the source, i.e. with the highest level of radiation. The collimating source channel must be open, i.e. created the worst radiation environment. By subtraction, find the technogenic component of the equivalent dose rate:

R.m - f

4. Under the same conditions, calculate the effective dose rate at the workplace. To do this, it is necessary to take into account the uneven irradiation of organs and tissues of the body near the source, i.e. measure DER T for 13 organs and tissues, and then multiply them by the weighting coefficients of radiosensitivity W T. Under our conditions, it is enough to confine ourselves to measurements for four control points of the body: 1 - head, 2 - chest, 3 - gonads, 4 - feet, and take for them, the enlarged weighting coefficients W K (see Table 2.1).

For the accepted position of the body at the workplace (“sitting” or “standing” as instructed by the teacher), measure the equivalent dose rate K at four control points. Subtract from all readings the average natural background f defined in clause 2.

= Σ ( K · W K), (2.1)

where k = 1…4 is the number of the control point of the body, K is the technogenic component of the DER and W K is the weighting coefficient of organs and tissues for each point (Table 2.1).

Table 2.1. To determine the effective dose rate at the workplace

Total: \u003d Σ ( K Wk) \u003d ___________ μSv / h

Find the radiation non-uniformity coefficient equal to the ratio of the effective dose to the readings of one dosimeter:

α = /

and to conclude whether it is expedient under the given conditions to take into account the unevenness of exposure when determining the effective dose.

6. Assuming that the student is at this workplace for all 16 hours of the laboratory workshop, determine the maximum possible effective dose of technogenic exposure of the student for the current year:

E stud = 16.

7. Based on the same considerations, estimate the maximum possible annual dose of group A personnel, assuming the standard working time of employees is 1700 hours:

E pers = 1700.

7. Determine the effective dose from natural exposure for the same calendar year (8760 hours), assuming that natural exposure affects human organs and tissues evenly:

E eat \u003d f 8760.

Estimate the possible spread of the dose of natural exposure, roughly accepting the confidence interval for the maximum and minimum background values ​​measured in paragraph 2.:

Δ = (max - min) 8760,

where max, min are background values. Present the value of the annual dose of natural exposure, taking into account the possible spread in the form E eat ± Δ/2 mSv.

8. Through an effective dose, evaluate the additional individual lifetime risk of non-threshold effects in students and employees, 1/(person · year) associated with the accepted working conditions:

r = E stud, Persian r E ,

where the risk coefficient is taken equal to r E = 5.6 10 – 2 1/ (person · · Sv).

9. Draw conclusions about the radiation safety in the laboratory, for which we compare the annual doses of technogenic exposure of employees and students with the corresponding dose limits of PD A and PD B. Calculate the factor of the margin to dose limits.

Compare the doses of technogenic exposure of employees and students with the expected annual dose from natural exposure and its dispersion.

2.3. Removal of dependence of dose rate on distance

In this part of the work, it is necessary to measure the dependence of the dose rate on the distance to the source using three different dosimeters in turn under conditions of an open and closed collimator on the container with the source.

With an open collimator a detector located in a gamma-ray beam "sees" a point source directly and registers its direct radiation. Absorption and scattering in air at short distances can be neglected, therefore, in this case, inverse square law: the intensity of radiation in vacuum is inversely proportional to the square of the distance from a point isotropic source, for example:

1 / 2 = (r 2 / r 1) 2 .

With a closed collimator The detector registers radiation significantly attenuated (by a factor of 300 or more) and scattered in the lead shield. The source of scattered radiation is the entire surface of the container, therefore, the source can no longer be considered a point source and the inverse square law can only be valid at large distances from it.

For measurements the detector of the selected dosimeter is mounted on a carriage that moves along a ruler with centimeter divisions. It is recommended to start from a far distance (r = 150 cm), and then, gradually bringing the detector closer to the source, find the boundary where the device does not go off scale. Take 4–5 dose rate readings at various distances in the selected range and subtract the background from them . Record the values ​​of distances and dose rates in the observation log (Table 2.2). Dosimeter readings should be converted into DER units (µSv/h) in the journal if the instrument is calibrated in other units.

Measurements should be repeated with several instruments with the collimator open and closed. At the same time, it should be taken into account that due to the different sensitivity of dosimeters, some of them can “go off scale” in the open beam, while others show nothing when closed. The UIM-2-2 device, calibrated in units of s –1, measures the photon flux through the detector (F) and is called radiometer. To convert its readings into dose rate units, you should use the calibration dependencies located on the desktop.

The results of measurements of DER dependence on distance should be presented on two graphs (one for an open collimator, the other for a closed collimator). On each of them, 3 curves are applied, smoothing the experimental points.

Table 2.2. Dose equivalent rate log

Note: from indications marked with *, the natural background should be subtracted.

2.4. Dose rate calculation from source activity

Dose rate calculations are conveniently performed in the form of Table. 2.3.

Table 2.3. Journal for dose rate calculations

BUT = ao/ 2n , (2.2)

where n is the number of half-lives that have passed from the date of metrological certification of the source to the date of the experiment: n = (t - To) / T 1/2

t is the current date of the experiment, To is the date of certification, T 1/2 is the half-life (n must be dimensionless); ao is the initial activity of the source according to the passport (data taken from the laboratory poster).

2. Recalculate in the same way on the date of the experiment the passport exposure dose rate at a distance of 1 m from the source, which is indicated on the laboratory poster on the date of its certification. Convert it to equivalent dose rate units (µSv/h).

3. Calculate DER values ​​at different distances from the source outside the protective container – o (r), µSv/h. For calculations, the inverse square law is used: the dose rate from a point isotropic source is directly proportional to its activity and inversely proportional to the square of the distance to it:

G · BUT/ r 2 , nGy /s, (2.3)

where is the absorbed dose rate, nGy/s; G is the gamma constant of the radionuclide, nGy × m 2 / (s × GBq); BUT is the source activity, GBq; r – distance, m.

To determine the equivalent dose rate (µSv/h), a radiation weighting factor W R is introduced into the formula, equal to one for gamma radiation, and a conversion factor 3.6 = 3600/1000:

O(r) = G BUT/ r 2 3.6 W R , µSv/h. (2.4)

Calculations according to the formula (2.4) should be written in the line with the number 2 of Table 2.3.

For the distance r =1 m, compare the DER value with the passport value obtained in step 2.

4. Make a correction for the attenuation of gamma radiation in the air. The thickness of the air layer is taken equal to the distance from the source to the detector, x = r.

The multiplicity of the weakening of the air layer with a thickness of x V cm is

K = exp (μ B x B) / B ∞ ,

where μ V is the linear coefficient of air attenuation, depending on the energy of gamma rays, cm–1; В ∞ is the accumulation factor in infinite geometry, which takes into account the contribution of radiation scattered by air (depends on the energy of gamma rays and on the product μх). These values ​​are taken according to tables A.1 and A.2 for the source gamma radiation energy.

DER at different distances, taking into account attenuation in air 1 = o / K, should be written in the 6th line of Table 2.3.

5. Calculate the DER values ​​at the same distances for the case when the source is in a closed lead container (the geometry of the lead shield can be considered barrier). The multiplicity of the weakening of lead protection with a thickness of x P b = 10.5 cm is

K R b \u003d exp (μ R b x R b) / (B R b d) ,

where μ R b is the linear attenuation coefficient of lead, taken from the energy of gamma rays (Table A.1); ВР b is the lead accumulation factor for infinite geometry, taken according to Table P.2, and d is the correction for the barrier geometry (depends only on the energy of gamma rays), taken according to Table P.3. DER taking into account attenuation in lead 2 = 1 / К Р b should be written in the 8th line of Table 2.3.

6. The results of calculations according to Table 2.3 should be plotted on two corresponding graphs obtained as a result of measuring DER from distance: one graph for the case of an unprotected source - 1 (r), the other for a source placed in a container - 2 (r). For convenience of reconciliation of dosimeter readings with calculations, experimental points from Table 2.2 should be shown on the graphs.

7. The conclusions on this part of the work should be:

Formulate the law of attenuation of radiation with increasing distance from the source;

think over possible reasons deviations of instrument readings from the calculated values;

Assess the absorbing capacity of air;

test questions

1. Effects of ionizing radiation on the human body.

2. Deterministic effects of radiation, mechanism of development.

3. Stochastic effects of radiation, development mechanism.

4. Direct and indirect effects of radiation on biological tissue.

5. Absorbed and equivalent dose - definition, units of measurement.

6. Effective dose, scope.

7. Collective dose and collective damage.

8. Dose rate. Natural radiation background.

9. Goals of radiation safety and ways to achieve them.

10. Principles of ensuring radiation safety.

11. The principle of justification.

12. The principle of regulation.

13. The principle of optimization.

14 Types of human exposure considered in NRB-99.

15. Types of radiation sources exempted from control and accounting.

16. Basic dose limits - definition and content of the concept.

17. Permissible levels for external technogenic exposure - connection with the main dose limits.

18. Gamma constant of the source. Relationship between the dose rate generated by a point isotropic source of γ-radiation, activity and distance.

19. The law of attenuation of radiation with distance.

20. The law of attenuation of radiation in matter.

21. Purpose, principle of operation and main characteristics of the devices used in this work. Possible areas of application of these devices.

22. Principles of protection against exposure to time, distance and screens.

23. Estimated exposure time and allowable dose rate.

24. Permissible operating time with a radiation source (when it should be evaluated and how).

Bibliographic list

2. Federal Law “On Radiation Safety of the Population”. No. 3-FZ dated 09.01.1996.

3. Norms radiation safety / NRB-99. - M.: TsSEN of the Ministry of Health of the Russian Federation, 1999. - 116 p.

4. Main sanitary rules for ensuring radiation safety / OSPORB-99. - M.: TsSEN of the Ministry of Health of the Russian Federation, 2000. - 132 p.

5. Kutkov, V.A. Basic provisions and requirements normative documents in the practice of ensuring the radiation safety of nuclear power plants: textbook / V.A. Kutkov [and others] - M: Izd. OIATE, 2002. - 292 p.

6. Kozlov, V.F. Reference book on radiation safety / V.F.Kozlov. – M.: Energoatomizdat, 1999. – 520 p.

7. Norms radiation safety NRB-76/87 and the Basic Sanitary Rules for Working with Radioactive Substances and Other Sources of Ionizing Radiation OSP-72/87 / USSR Ministry of Health. – M.: Energoatomizdat, 1988. – 160 p.

8. Golubev, B.P. Dosimetry and protection from ionizing radiation / B.P. Golubev. – M.: Energoatomizdat, 1986. – 464 p.

Application

Table A.1. Linear attenuation coefficients μ , cm–1, for some substances depending on the energy of photon radiation

Table A.2. Dose accumulation factors in infinite geometry B ∞

for a point isotropic source

Table A.3. Amendment to Table A.2 for calculating the accumulation factor AT b point isotropic source in barrier geometry ( d = B b/c ∞ )

1. BASICS OF RADIATION SAFETY……………….…………....3

1.1. Biological effect of ionizing radiation………………….……..3

1.2. Threshold and non-threshold effects in human exposure…….…….…5

1.3. Basic dosimetric quantities and units of their measurement……………………………………………………………………………..12

1.4. Basic provisions of the NRB-99 Radiation Safety Standards……..…15

2.1. Preparing for work……………………………………………………….….18

2.2. Assessment of radiation safety when working with a source……….….19

2.3. Removal of dependence of dose rate on distance………………………..21

2.4. Calculation of dose rate by source activity…………………………..23

Control questions……………………………………………………………..25

Bibliographic list…………………………………………………….…26

Application………………………………………………………………………..26

International Commission on Radiological Protection, established in 1928. at the 2nd International Radiological Congress. Together with the International Commission on Radiation Units and Measurements (ICRU, 1925), brings together experts in the field of radiation measurements, the biological effects of radiation, dosimetry and radiation safety.

United Nations Scientific Committee on the Effects of Atomic Radiation. Established by the UN in 1955 to assess the health effects of exposure to ionizing radiation.

To reduce the impact of external gamma radiation, three main methods are used worldwide:

Time;
Distance;
Shielding (installation of protection).

Time

DOSE = DOSE RATE * TIME

One of the factors affecting the radiation dose is time.

The dependency is simple: less time effects of AI on the body - less dose.

A rough estimate can help determine the dose a worker will receive over a period of time, or how long they can stay at work without reducing the dose rate.

For example:

The worker is going to do a job that takes about an hour and a half. The dose rate at the workplace is 1.0 mSv/h (mSv/h). Determine the expected radiation dose.

DOSE = DOSE RATE * TIME = 1.0 mSv/h (mSv/h) * 1.5 h (h) = 1.5 mSv (mSv).

Answer: the committed dose would be 1.5 mSv (mSv).

If the worker works faster and finishes the job in one hour, then he will reduce the dose to 1.0 mSv (mSv): (1.0 mSv/h * 1.0 h = 1.0 mSv).

If a break from work is necessary (for rest, etc.), then the worker should move out of the AI ​​impact area to a place where the radiation level is as low as possible.

Distance

Based on the formula for calculating the radiation dose:

DOSE = DOSE RATE * TIME

Low dose rate means a small dose of radiation. A property of all AI sources is that the dose rate decreases with distance.

The radiation source can have a different configuration: point, volume, surface or line source.

Radiation from a point source decreases with the square of the distance. For example:

The dose rate at a distance of one meter from the source is - 9 mSv/h (mSv/h). If the worker increases the distance to three meters, the dose rate will be reduced to 1 mSv/h (mSv/h).

However, most radiation sources are not point sources. There are a lot of linear sources, there are also large volumetric sources such as radioactive tanks and heat exchangers.

For line sources and large sources, the dose rate decreases in proportion to the distance.

At a distance of one meter from the source, the dose rate is 9 mSv/h (mSv/h). At a distance of three meters, it will be - 3 mSv / h (mSv / h).

As the distance from the IR source increases, the dose rate will also decrease.

Simple and effective measure protection from AI - to be as far away from the source of ionizing radiation as possible.

Protection (shielding)

Based on the formula for calculating the radiation dose:

DOSE = DOSE RATE * TIME

As mentioned above, the dose rate to which a worker is exposed determines the dose of radiation that he receives. The lower the dose rate, the lower the radiation dose.

The dose rate can be reduced by installing protection (shielding), since any matter absorbs radiant energy when irradiated. That is why the worker is exposed to less radiation if there is protection between him and the radiation source.

Pay attention to alpha, beta, and gamma radiation affecting thin sheet of paper. As you know, the range of alpha radiation is quite short. It stops with a thin layer of skin, especially a sheet of paper. Beta and gamma radiation will not stop a sheet of paper.

Plexiglass(see figure 7.8) will stop the beta emission completely. Gamma rays will be somewhat attenuated, but generally pass through plexiglass freely.

The next type of protection is a lead protective screen. Here the gamma radiation will be reduced, but it will not be completely stopped.

Gamma radiation, the most common type of radiation in a nuclear power plant, cannot be completely shielded, it can only be reduced. The best shielding materials are concrete and water.

The optimal thickness of the protective screen depends on the energy of the radiation and the activity of the radiation source. Calculating the thickness of the protection is quite complicated, but you can use the "rule of thumb".
1 centimeter of lead will reduce the dose rate of gamma radiation (cobalt-60) by half.
5 centimeters of concrete will reduce the dose rate of gamma radiation (cobalt-60) by half.
10 centimeters of water will reduce the dose rate of gamma radiation (cobalt-60) by half.

The placement and removal of protective screens is carried out with the permission and under the guidance of the service of the Republic of Belarus!