Calculation of protection against ionizing radiation. Open Library - open library of educational information

Sanitary rules for the design and operation of radiation circuits in nuclear reactors*

APPROVED by Deputy Chief State Sanitary Doctor of the USSR A.I. Zaichenko on December 27, 1973 N 1137-73
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* These Rules were developed by employees of the branch of the Scientific Research Institute of Physics and Chemistry named after. L.Ya. Karpov and the All-Union Central Research Institute of Labor Protection of the All-Union Central Council of Trade Unions.

Introduction

Introduction

These rules are drawn up to develop the “Radiation Safety Standards”* (NRB-69) and the “Basic Sanitary Rules for Working with Radioactive Substances and Other Sources of Ionizing Radiation”* (OSP-72).
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SP 2.6.1.2612-10 (OSPORB-99/2010);
** The document is not valid on the territory of the Russian Federation. SanPiN 2.6.1.2523-09 (NRB-99/2009) is in force. - Database manufacturer's note.

The rules are mandatory for all institutions and enterprises designing, constructing and operating radiation circuits (RC) at nuclear reactors.

The rules apply to RKs of research, semi-industrial and industrial types, intended for carrying out radiochemical processes, radiation sterilization, biological experiments, etc.

Responsibility for the implementation of these Rules rests with the administration of institutions (enterprises).

1. Basic concepts, definitions and terminology

1.1. Radiation circuit (RC) is a device for gamma irradiation that uses the circulation of working substances in which gamma-active isotopes are formed under the influence of reactor neutrons.

1.2. Gamma carrier is a working substance that is a source of gamma radiation in the Republic of Kazakhstan.

1.3. A fissile gamma carrier is a substance in which atomic nuclei are split under the influence of neutrons.

1.4. An activity generator is a device in which the working substance RK becomes gamma-active.

1.5. Irradiator is a part of the RK intended for irradiating various objects with gamma carrier radiation.

1.6. A radiation apparatus is a device designed to carry out a specific radiation process.

1.7. Delayed neutrons are neutrons emitted by nuclei some time after fission.

1.8. Photoneutrons are neutrons emitted from atomic nuclei as a result of their interaction with gamma rays.

1.9. RCs with a water method of protection are those RCs in which the irradiator is constantly under a protective layer of water.

1.10. RKs with a dry method of protection are those RKs in which concrete, lead and other solid materials are used for radiation protection.

1.11. A working chamber is a room surrounded by protection in which irradiation is carried out.

1.12. Working pool - a pool used to store the irradiator and to house the irradiated object.

1.13. A labyrinth (curved corridor) is a typical protective device that protects against radiation from a source outside the working chamber.

1.14. Gamma carrier storage is a special container connected to the RK system in which the gamma carrier is stored when circulation stops.

1.15. Emergency storage is a special container (reservoir) designed to drain gamma carriers in emergency situations.

1.16. The operator's room is the room in which the radio control systems are located.

1.17. Adjacent room - a room directly adjacent to the working chamber and separated from it by a permanent partition (wall, floor, ceiling).

1.18. The prohibited period is the operating time of ventilation after the end of irradiation, necessary to reduce the concentration of toxic substances in the working chamber to the maximum permissible values.

2. General provisions

2.1. According to their purpose, the control agents for nuclear reactors are divided into two groups:

Group I - RK of scientific research, semi-industrial and industrial types, intended for carrying out explosive processes;

Group II - RK of scientific research, semi-industrial and industrial types, intended for carrying out non-explosive processes.

2.2. When developing RKs and their operation, the specific features of the type of reactor used and the properties of the gamma carrier used must be taken into account.

2.3. The degree of possible radiation hazard during the operation of the Republic of Kazakhstan is determined by the following main factors:

a) the intensity of external gamma radiation fluxes in work areas;

b) radioactive contamination of premises, equipment and irradiated objects resulting from depressurization of the RK system and during repair work;

c) air pollution production premises radioactive aerosols and gases;

d) the intensity of delayed neutron fluxes when using a gamma carrier on fissile materials;

e) the intensity of photoneutron fluxes generated by the reaction (, );

f) activation of irradiated objects, radiation devices, and the environment by delayed neutrons and photoneutrons.

2.4. Non-radiation sources of danger are:

a) ozone and nitrogen oxides formed as a result of air radiolysis;

b) products of radiolysis of water, if present in technological systems of the Republic of Kazakhstan;

c) toxic substances entering indoor air from irradiated objects, etc.

2.5. Potential hazards include:

a) explosive and flammable substances irradiated at the RK, or products formed during the irradiation process;

b) an “explosive mixture”, the formation of which is possible during radiolysis of water in the case of placing individual RK units under water;

c) aggressive environments that arise during the operation of the Republic of Kazakhstan.

2.6. Projects of newly built during* reconstruction of the Republic of Kazakhstan are subject to mandatory approval by the sanitary and epidemiological service institutions. RoK projects must take into account all hazard factors and develop effective measures to reduce harmful effects on personnel.
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* The text of the document corresponds to the original. - Database manufacturer's note.

2.7. Before they are put into operation, the RK must be accepted by a commission consisting of representatives of the administration of the institution (enterprise), the sanitary and epidemiological service, Gosatomnadzor and other interested organizations.

2.8. Persons who do not have medical contraindications listed in the appendix to the "Basic Sanitary Rules". A medical examination should be carried out once a year, and monitoring of the content of radioactive substances in the body of workers during trouble-free operation of the Republic of Kazakhstan - once every 5 years.

2.9. Based on these Rules, the administration of the institution (enterprise) develops detailed safety instructions for servicing and working on the RC, taking into account the features of the RC structure and the work being carried out.

2.10. Responsibility for the safety of work in the Republic of Kazakhstan lies with the administration of institutions (enterprises) and work managers.

2.11. Everyone working in the Republic of Kazakhstan must be trained in safe work methods, know the rules for using sanitary devices, protective equipment and personal hygiene rules, and also pass the appropriate technical minimum. Repeated knowledge testing should be carried out at least once a year. Persons involved in work in the Republic of Kazakhstan must be instructed before starting work. In case of changes in a number of RK parameters (irradiation process technology, RK control system, etc.), it is necessary to conduct additional instructions.

3. Requirements for the design and protection of radiation circuits

3.1. RCs with gamma carriers of any type must have a reliable sealing system.

3.2. Materials used for the manufacture of components and communications of the Republic of Kazakhstan must have:

a) sufficient mechanical strength;

b) high corrosion resistance under operating conditions;

c) low sorption capacity in relation to the gamma carrier;

d) low activation cross section in neutron fluxes;

e) short half-life of the induced activity.

3.3. The most vulnerable components and systems of the circulation system (electromagnetic pumps, level sensors, temperature sensors, etc.) must be located in such a way that their replacement, in the event of failure, is carried out with minimal danger and without violating the tightness of the circulation system.

3.4. When designing a receptacle, it is advisable to choose, under other conditions, the lowest speed of circulation of the gamma carrier to reduce corrosion and erosion of the structural materials of the dispenser.

In the case of using fissile material as a gamma carrier, the circulation rate must ensure, in addition, the minimum activity induced by delayed neutrons in the irradiated system and the structural materials of the reactor.

3.5. The design of the reactor must provide for the prevention of blockages in communication systems under any operating modes of the nuclear reactor.

When designing a distribution system based on the calculation of the thermal conditions of all nodes and communications of the distribution system, the possibility of such blockage must be excluded. The design of the RK should provide for the possibility of eliminating the blockage of communications by a gamma carrier.

During the operation of the RK, it is necessary to constantly monitor the temperature of the gamma carrier and, if necessary, take measures to maintain the operating mode.

3.6. The design of the RK should allow the gamma carrier to be completely removed, if necessary, into a special storage facility (drainage device, etc.). It is necessary to ensure such an arrangement of the RK nodes and communications and such a design of the irradiator that maximally facilitate the natural removal of the gamma carrier into the storage facility. In this case, it is necessary to take into account the change in reactor power due to the emergency discharge of the gamma carrier.

3.7. The RC must be equipped with a device for the forced removal of gamma carrier residues into a special storage facility (for example, by purging the RC system with inert gases, etc.), as well as the removal of gamma carrier from those components of the RC from which it cannot be discharged under the influence of gravity.

3.8. When accepting the RC into operation, after eliminating detected installation defects, the circuit is loaded with gamma carrier and the reliability and stability of its circulation is checked both in starting and in stationary circulating modes (the first stage of acceptance). In the second stage of acceptance, during the circulation of the gamma carrier at low power of the nuclear reactor (close to zero), the reliability and stability of all RK systems, including dosimetric and technological control devices, are checked. At the final stage of acceptance, the commission checks the amount of gamma background at the outer surfaces of the protection in the process of gradually bringing the reactor to maximum power.

At the final stage, the commission draws up an act on acceptance of the Republic of Kazakhstan for operation.

3.9. The calculation of RK protection should be carried out taking into account all types of radiation (neutrons, gamma radiation, etc.).

3.10. When using non-fissile gamma carriers in the Republic of Kazakhstan, the calculation of protection is carried out according to the universal tables given in Appendix 1.

4. Requirements for interlocking and alarm systems

4.1. RKs must have reliable blocking and alarm systems that provide continuous information about radiation levels and are triggered independently of each other both when the dose rate increases and when technological systems malfunction. RKs with dry-type protection must be equipped with at least two completely independent locking systems for the entrance door of the irradiation chamber (or labyrinth).

4.2. If at least one of the locking and alarm systems of the entrance door of the irradiation chamber is malfunctioning, the operation of the RK is prohibited until the malfunction is eliminated.

4.3. Locking systems should be based on the simultaneous use of:

a) devices that inform about the dose rate of gamma and neutron radiation;

b) a device (pump, etc.) that ensures circulation of the gamma carrier in the RK system.

4.4. When the front door is unlocked, the gamma carrier must be kept in storage, and the possibility of its circulation must be excluded.

The possibility of a person getting into the working chamber and labyrinth in the case of a conveyor system for supplying objects for irradiation during RK operation should also be excluded.

4.5. The front door must remain locked when the power is turned on.

4.6. The working chamber of the RK must be equipped with a sound and light alarm, which warns of the need to immediately leave the working chamber (or labyrinth).

4.7. Entry into the working chamber of the Republic of Kazakhstan is allowed only with the permission of the responsible person on duty.

4.8. The working chamber (or labyrinth) must contain devices that make it possible to immediately stop the circulation of the gamma carrier and transfer it to storage.

4.9. The RK control panel must have instruments and a light display informing about the dose rates of gamma and neutron radiation (for a circuit with fissile material) in the working chamber, in the labyrinth, about the operation of devices for circulating the gamma carrier, vacuum systems, etc. It is necessary to equip the RK with sensors that signal the leakage of gamma carrier from the circuit.

4.10. If a prohibited time period is established, the entry door lock must include a device to ensure that the time period is respected after the gamma carrier has been removed.

4.11. On conveyor belts equipped with a conveyor and installation hatches, the possibility of people getting into the working chamber through the inlet and outlet openings of the conveyor and opening the hatch during operation of the conveyor must be excluded.

4.12. RKs with water protection must be equipped with sound and light alarms:

a) about changes in water level;

b) about increasing the threshold dose rate above the pool water surface.

4.13. When the water level in the pool decreases, leading to an increase in the level of radiation exceeding that provided for this installation, an autonomous blocking system must ensure that the circulation of the gamma carrier is stopped and transferred to storage.

4.14. The pool must have fences or a cover to prevent accidents during repairs and other work on the RoK.

5. Ventilation requirements

5.1. Ventilation of premises in the Republic of Kazakhstan is designed taking into account the requirements of SN-245-71* and must ensure the removal, along with radioactive aerosols and gases, of air radiolysis products and other toxic substances released or formed from irradiated materials and equipment.
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* The document is not valid on the territory of the Russian Federation. SP 2.2.1.1312-03 applies, hereinafter in the text. - Database manufacturer's note.

5.2. In all rooms where RK communications pass, it is necessary to create a vacuum of about 5 mm of water column, which ensures air suction from clean rooms. Ventilation ducts of exhaust ventilation systems must be made of materials that are resistant to corrosion and do not absorb radioactive substances.

5.3. The working chamber must be equipped with supply and exhaust ventilation with an excess of exhaust over inflow by 10-15%. In winter, it is necessary to provide heating of the supplied air. The working chamber and the control room must be served by independent ventilation systems with separate air ducts and fans running constantly. It is allowed to turn off the fans while the gamma carrier is in the storage.

5.4. The frequency of air exchange required to reduce air pollution by radioactive and toxic substances to values ​​not exceeding the average annual permissible concentration (AAC) is calculated depending on the gamma power of the RK and the volume of the working chamber. In cases where, for one reason or another, the required air exchange rate cannot be ensured, a prohibited period of time is introduced.

5.5. The RK control panel must be equipped with a sound and light alarm system to notify of malfunctions or stopping of the fans.

5.6. The ventilation system must ensure the purification of the air from radioactive aerosols and gases in the event of an emergency release.

6. Requirements for the premises of the Republic of Kazakhstan and means of eliminating radioactive contamination

6.1. Depending on the characteristics of the RC device and the conditions of its operation, when planning premises, it is necessary to provide for a clear delineation of rooms where contamination is possible due to depressurization of RC communications and from other rooms with equipment on their boundaries of devices for personal protective equipment.

6.2. The walls, ceiling of the working chamber, temporary storage rooms for radioactive waste, as well as all working surfaces and equipment are covered with low-absorbing, easily decontaminated materials that are resistant to gamma carriers.

6.3. When designing a reactor system in a nuclear reactor complex, the following must be provided:

devices for checking the tightness of the RK system;

room for temporary storage of radioactive waste.

6.4. In the working chamber or in the adjacent room, devices must be provided to eliminate radioactive contamination in the event of depressurization of the radioactive system, decontamination systems and special sewage systems must be equipped.

In the event of radioactive contamination caused by a gamma carrier, the operation of the RK is prohibited until the causes are clarified and the accident is eliminated.

6.5. It is advisable to make all communications from seamless pipes and with a minimum number of welded and other connections. The places where the RK communications pass through the reactor basin and the structures (protection, partition, etc.) separating the reactor core from the RK working chamber must be sealed with the obligatory preservation of the “pipe-in-pipe” principle.

7. Radiation and preventive control

7.1. Dosimetric monitoring in the Republic of Kazakhstan, as well as monitoring of compliance by all workers with the requirements of these Rules is carried out by the radiation safety service of this institution (enterprise).

7.2. The Radiation Safety Service carries out:

a) control of individual doses of external radiation;

b) control of external exposure levels in workplaces and adjacent rooms;

c) control over contamination of the working surfaces of equipment and irradiated objects, clothing, shoes and skin of service personnel;

d) control of radioactive contamination of water in the pool;

e) control over the content of radioactive gases and aerosols.

7.3. Monitoring the efficiency of fans and the content of toxic substances in the air is carried out by a special service of the enterprise (organization).

7.4. In cases where activation of irradiated objects by neutrons is possible, it is also necessary to control their induced activity.

7.5. Individual cards are issued for all persons working in the Republic of Kazakhstan, in which monthly and annual doses of external radiation are entered.

7.6. The frequency of radiometric and dosimetric measurements and the nature of the necessary measurements are established by the administration of institutions (enterprises) in agreement with local sanitary and epidemiological service authorities.

7.7. All repair, maintenance and emergency work must be carried out under radiation monitoring using personal protective equipment. The set of personal protective equipment and the permissible time for carrying out work are determined by the radiation safety service.

7.8. Technical projects must provide for systems for stationary monitoring of the Republic of Kazakhstan and equipping the radiation safety service with modern equipment necessary to carry out appropriate measurements and analyzes, taking into account the characteristics of gamma carriers and irradiated objects.

8. Accident prevention measures

8.1. All manipulations with the irradiator and communication systems of the Republic of Kazakhstan must be carried out in such a way as to prevent their mechanical damage.

8.2. If the normal operation of the RK is disrupted (for example, temperature deviation from the specified operating intervals, etc.), the gamma carrier must be removed to storage.

8.3. When developing a device intended for circulation of gamma carrier, it is necessary to provide methods to prevent hydraulic shocks of liquid gamma carrier in the communications system of the Republic of Kazakhstan.

8.4. In RK projects with a water cooling method for RK systems, measures must be taken to prevent the formation of an explosive concentration of an explosive mixture.

8.5. In Group II RK, irradiation of explosive substances is permitted in special cylinders that are known to be able to withstand an explosion of the irradiated substance

8.6. When carrying out the process of loading toxic gamma carriers into the Republic of Kazakhstan, as well as when carrying out repair, preventive and emergency work, it is necessary to use individual means protections preventing the entry of these substances and compounds into skin and into the body of workers (taking into account the toxicity of the gamma carrier).

8.7. For Group I RC, the following must be provided:

a) automatic, duplicating each other systems, which, in the event of a threat of explosion (for example, an increase in temperature or pressure in the irradiated object above the permissible level), allow the gamma carrier to be immediately transferred to the storage position;

b) the design of the radiation apparatus in which the explosive substance is irradiated, ensuring the integrity of the irradiator and communication systems in the event of an explosion;

c) the design of the protection of the working chamber, which must be such that it does not collapse in the event of an explosion; The entrance to the working chamber must be protected by a blast door.

8.8. To carry out explosive radiation processes, the use of radioactive materials with a fissile gamma carrier, as well as with a gamma carrier with a half-life of more than 100 hours, is undesirable.

8.9. In the event of an explosion in the Republic of Kazakhstan, which caused damage to the irradiator and communication systems and led to contamination of the working chamber with gamma carriers, entry into it is allowed only after a certain time of exposure of the gamma carrier with the permission of the radiation safety service.

8.10. The organization's radiation safety service must develop detailed instructions in case of emergency situations, taking into account the specifics of the design of the RK and the ongoing radiation processes, indicating the necessary measures to eliminate accidents.

These Rules apply to all designed, constructed and operating control systems for nuclear reactors and come into force from the moment of their publication. Previously existing Rules for the Republic of Kazakhstan No. 654-66 are cancelled.

In cases where large capital expenditures are required to re-equip existing RCs in accordance with the requirements of these Rules, the issue of such re-equipment is resolved in each case separately in agreement with the local sanitary and epidemiological service authorities.

Appendix 1. Calculation of protection from gamma radiation of radioactive isotopes K_(42), In_(116m), Mn_(56) and Na_(24)

Annex 1

Calculation of protection against gamma radiation of radioactive isotopes K, In, Mn and Na

To determine the required thickness of protection from the tables, there are two input arguments: the top horizontal line shows the radioactive isotopes K, In, Mn and Na for four protective materials (water, concrete, iron and lead), the left vertical column shows the attenuation factor, the remaining columns contain the required protection thickness (cm) for the corresponding material and gamma carrier. The following material densities are accepted: for water - 1.0 g/cm, for concrete - 2.3 g/cm, for iron - 7.89 g/cm, for lead - 11.34 g/cm.

The tables for attenuation factors are compiled in sufficient detail, so that for intermediate values ​​the protection thickness can be found by simple linear interpolation. If a factor of attenuation of more than 10 is required in the calculations, then it is permissible to extrapolate the thicknesses based on the comparative effect of the last tabulated factors of attenuation. The tables can be applied not only to point sources, but also to extended sources.

Examples of calculation of protection based on dose rate reduction factors

Accepted designations: - total activity, expressed in milligram equivalents of radium, - distance from the source in meters, - thickness of protection in centimeters, - dose rate in µR/s at the workplace without protection, - maximum permissible level of dose rate at the workplace , microdistrict/s.

If the values ​​of and are known, then the required attenuation factor is found by the formula:

If the source activity is specified in mEq of radium and the distance from the source to the workplace in centimeters, the dose rate (μR/s) can be calculated using the formula:

Similar to the previous case.

Based on the value found (left vertical column), the thickness of the protection for the corresponding material and gamma carrier is found.

Example 1.

The measured or calculated dose rate at the workplace is given as 1.55 r/s. The source of -radiation is In. Find the thickness of the concrete screen required to attenuate this radiation to the maximum permissible value of 1.4 mr/h.

Solution:

Attenuation factor. From the tables we find that for the In and 4 10 isotopes the thickness of the protection is 159 cm.

Example 2.

The source of radioactive sodium (Na) has an activity of 200 g-equiv of radium and is located in the irradiator of a radiation chemical installation. Find the thickness of the lead wall separating the control panel from the source, if 10 m and the dose rate should be reduced to the level of 0.4 µR/s.

Solution:

The dose rate from an unprotected source for 10 m is equal to: µR/s.

Attenuation factor.

The required thickness for Na is 17.5 cm.

Calculation of protection from -rays of a circulating mixture of unseparated fission fragments (radiation circuits with fissile material) must be carried out individually for each specific case, since at present it is impossible to provide compact tables for such calculations.

Select the cross-section of the traverse beam and rope for lifting the rolling mill spindle.

Initial data:

Spindle weight Q=160 kN;

traverse length l=6m;

the traverse beam bends.

Draw up a slinging diagram.

Select the section of the cross beam, the type and section of the rope.

Solution:

Slinging scheme with a traverse at two points.

Rice. 21 – Slinging diagram. 1 – center of gravity of the load;

2 – traverse; 3 – roller; 4 – sling

Determination of the tension force in one branch of the sling

S = Q / (m cos) = k Q / m = 1.42 160 / 2 = 113.6 kN.

where S is the design force applied to the sling without taking into account overload, kN;

Q – weight of the lifted load, kN;

 – the angle between the direction of action of the design force of the sling;

k – coefficient, depending on the angle of inclination of the sling branch to the vertical (at =45 o k=1.42);

m – total number of sling branches.

Determine the breaking force in the sling branch:

R = S · k з = 113.6 · 6 = 681.6 kN.

where k з is the safety factor for the sling.

We choose a rope of type TK 6x37 with a diameter of 38 mm. With a calculated tensile strength of the wire of 1700 MPa, having a breaking force of 704,000 N, i.e., the closest greater to the breaking force required by calculation of 681,600 N.

Selection of cross-beam cross-section

Fig. 22 – Design diagram of the traverse

P = Q k p k d = 160 1.1 1.2 = 211.2

where k p is the overload coefficient, k d is the load dynamic coefficient.

Maximum bending moment in traverse:

M max = P a / 2 = 211.2 300 / 2 = 31680 kN cm,

where a is the traverse arm (300 cm).

Required moment of resistance of the cross-section of the cross-beam:

W tr > = M max / (n R from ) = 31680 / (0.85 21 0.9) = 1971.99 cm 3

where n = 0.85 – working conditions coefficient;

 – bending stability coefficient;

R from – design resistance to bending in the traverse, Pa.

We select the design of the traverse beam with a through section, consisting of two I-beams connected by steel plates No. 45 and determine the moment of resistance of the traverse as a whole:

W d x = 1231 cm 3

W x = 2 · W d x = 2 · 1231 = 2462 cm 3 > W tr = 1971.99 cm 3,

which satisfies the strength condition for the design cross-section of the traverse.

9. Structural and strength calculations

9.1. Calculation of the protective casing of a multi-spindle vertical semi-automatic lathe Example 37

Initial data:

The protective casing of a multi-spindle vertical semi-automatic lathe is a rectangular steel structure with a length of l = 750 mm, a width of b = 500 mm and a thickness of S. It is clamped in holders at the ends so that the system can be considered as a beam lying on two supports.

The chips have a weight G = 0.2 g and fly towards the casing at a speed of V = 10 m/s and strike the casing perpendicularly to its middle.

Distance from the place of chip separation in the cutting zone to the casing:

Determine the thickness of the sheet from which the protective casing can be made.

SOLUTION:

As a result of the impact of the chips, the casing becomes deflected. The greatest deflection will be caused by chips caught in its middle. The pressure that corresponds to this deflection is:

,

where E is the elastic modulus of the casing material. For steel sheet:

E = 2·10 6 kg/cm2;

I – moment of inertia of the beam – casing. For a rectangular section:

f – deflection of the casing at the point of impact:

l – casing length.

The energy accumulated in the casing is equal to:

At the moment of maximum deflection of the casing, the force will be converted entirely into the potential energy of deformation of the casing, i.e.

Calculation of protection against alpha and beta radiation

Time protection method.

Distance protection method;

Barrier (material) protection method;

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

To calculate 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 density of the absorbing substance.

Protection from alpha radiation at nuclear power plants (for example, when receiving “fresh” fuel) due to the short path lengths in the substance is not difficult. Alpha-active nuclides pose the main danger only during internal irradiation of the body.

The maximum free path of beta particles can be determined using 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 practice at nuclear power plants, there are gamma radiation sources of various configurations and sizes. The dose rate from them can be measured with appropriate instruments or calculated mathematically. In general, 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 a point source . It represents a gamma emitter for which, without a significant loss of calculation accuracy, its dimensions and self-absorption of radiation in it can be neglected. In practice, any equipment that is a gamma emitter 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 attenuation factor K of radiation and the energy of gamma rays. Such tables are given in reference books on radiation safety and are calculated based on the formula for the attenuation in matter of a wide beam of photons from a point source, taking into account the accumulation factor.

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

Rice. 6.1. Diagram of a narrow photon beam

1 - container; 2 - radiation source; 3 - diaphragm; 4 - 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 shield as a result of its interaction with matter occurs according to an exponential law:

I = I 0 e - m x (6.1)

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

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

Expression (7.1) is also valid when using the mass attenuation coefficient m m instead of the linear one. In this case, the thickness of the protection should 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 measurement results of narrow and wide beams is characterized by the accumulation factor B:

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

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

where m, m shir is the linear attenuation coefficient for narrow and wide photon beams, respectively. Values ​​of m and IN for various energies and materials are given in radiation safety reference books. 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 for protection against photon radiation: lead, steel, concrete, lead glass, water, etc.

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

P av = PD A /1700 hour = 20 mSv / 1700 hour = 12 μSv/hour;

where P av – permissible level of average annual effective (equivalent) dose rate;

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

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

K = P meas / P avg;

where Rmeas 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) using universal tables, one should know the photon energy e (MeV) and the radiation attenuation factor K .

In the absence of universal tables, a quick determination of the approximate thickness of the protection can be performed using approximate values ​​of the photon half-attenuation in the wide beam geometry. The half-attenuation layer Δ 1/2 is a protection thickness 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 factor and the number of half-attenuation layers:

With a known number of half-attenuation layers n, the thickness of the protection is 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.

Method of protection by distance. The dose rate of photon radiation from a point source in a void 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 Px at any other distance Rx is calculated by the formula:

P x = P 1 R 1 2 / R 2 x (6.4)

Time protection method. The time protection method (limiting the time a worker spends under the influence of ionizing radiation) is most widely used when performing radiation-hazardous work in a controlled access zone (CAZ). These works are documented in a dosimetry work order, which indicates the permitted time for the work.

Chapter 7 METHODS OF REGISTRATION OF IONIZING RADIATION

Technical means of protection include the construction of various screens made of materials that reflect and absorb radioactive radiation. Screens are installed 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 attenuation factor. The characteristics of protective materials and experience with radiation sources make it possible to outline the preferential areas of use of a particular 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 these requirements well: lead glass, lime glass, glass with liquid filler (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 types radiation with matter. Protection from 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 of a few centimeters or a sheet of paper is sufficient protection against alpha particles.

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

Lead shields are used to protect against high energy beta particles (electrons), but the inner lining of the shields must be made of a material with a low atomic number to reduce the initial energy of the electrons, and therefore the energy of the radiation 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, it is first necessary to take into account the spectral composition of the radiation, its intensity, as well as the distance from the source where the operating personnel are located, and the time spent in the area of ​​exposure to radiation.

Currently, based on the available calculated and experimental data, tables of the attenuation factor are known, as well as various types of nomograms that make it possible to determine the thickness of protection against gamma radiation of various energies. As an example in Fig. 59 shows a nomogram for calculating the thickness of lead protection from a point source for a wide beam of Co60 gamma radiation, which ensures a reduction in the radiation dose to the maximum permissible. The abscissa axis shows the protection thickness d, the ordinate axis shows the coefficient K1 equal to

(24)

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

t is the operating time in the area of ​​radiation exposure, h; R is the distance from the source, cm. For example, it is necessary to calculate the protection from the Co60 source, at M = 5000 mEq Ra, if the operating personnel is 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 ordinate axis is the attenuation factor K, equal to

K=D0/D

Using expression (23), we obtain

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

Rice. 59. Nomogram for calculating the thickness of lead protection 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 a gamma-therapeutic unit is located, charged with a Cs137 drug of 400 g-eq Ra (M = 400,000 mg-eq Ra). The closest distance at which maintenance personnel are located in the adjacent room is R = 600 cm. According to sanitary standards, in adjacent rooms in which there are people not involved in working with radioactive substances, the radiation dose should not exceed 0.03 rem/week or for gamma radiation is approximately 0.005 rad per working day, i.e. D = 0.005 rad per t = 6 hours of attenuation, we use formula (23). To estimate 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, you must be guided by its structural properties, as well as the requirements for the size and weight of the protection. For protective casings of various types (gamma therapeutic, gamma flaw detection), when mass plays a significant role, the most advantageous protective materials are those that best attenuate gamma radiation. The greater the density and atomic number of a substance, the more degree attenuation of gamma radiation.

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

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

When creating stationary protection (i.e. protecting rooms in which work is carried out with gamma sources), ensuring the stay of people in adjacent 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 high serial number, in particular barite, are added to the 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 immersed in a pool of water, the thickness of the layer of which ensures the necessary reduction in the radiation dose to safe levels. If there is water protection, it is more convenient to charge and recharge the unit, as well as carry out repair work.

In some cases, working conditions with gamma radiation sources may be such that it is impossible to create stationary protection (when recharging installations, removing a radioactive drug from a container, calibrating a device, etc.). What is meant here is that the activity of the sources is low. To protect service personnel from radiation 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 carried out using long grips or holders. In addition, this or that operation must be performed only during that 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 supervision of a dosimetrist. At the same time, there should be no unauthorized persons in the room, and the area in which the dose exceeds the maximum permissible during work must be fenced off.

It is necessary to periodically monitor the protection using dosimetric instruments, since over time it may partially lose its protective properties due to the appearance of certain unnoticeable 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 using the appropriate formulas or nomograms. In this case, substances with a low atomic number should be taken as protective materials, because with each collision with a nucleus, the neutron loses the greater part 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 fluxes of gamma radiation, which are formed during the fission process, as well as during the decay of fission products. Therefore, when designing protection against neutrons, it is always necessary to simultaneously provide protection against gamma radiation.

Helpful information:

Federal Agency for Education

State educational institution

higher vocational education

"Ivanovo State Energy University

named after V.I. Lenin"

Department of Nuclear Power Plants

RADIATION SAFETY
AND DOSIMETRY OF EXTERNAL GAMMA RADIATION

Guidelines for performing 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” who are undergoing a laboratory workshop on the physics of ionizing radiation. The theoretical material presented in section 1 complements and partially duplicates what is read in lectures.

Approved by the cycle methodological commission of the IFF

Reviewer:

Department of Nuclear Power Plants, State Educational Institution of Higher Professional Education "Ivanovo State Energy University named after V. I. Lenin"

RADIATION SAFETY AND DOSIMETRY

EXTERNAL GAMMA RADIATION

Guidelines for laboratory work No. 1

on the course “Radiation Protection”

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.

Printing is flat. Conditional oven l. 1.62. Circulation 100 copies. Order No.

GOUVPO "Ivanovo State Energy University named after V. I. Lenin"

153003, Ivanovo, st. Rabfakovskaya, 34.

Printed in UIUNL ISUE

1. FUNDAMENTALS OF RADIATION SAFETY

1.1. Biological effects of ionizing radiation

Ionizing radiation, affecting a living organism, causes in it a chain of reversible and irreversible changes, the “trigger mechanism” 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 an ionizing particle (α, β - particle, X-ray or γ - photon) transfers to the electron shell of the atom energy sufficient to remove an orbital electron (i.e. exceeding binding energy). If the transferred part of the energy is less than the binding energy, then only excitation occurs electron shell atom.

In simple substances, the molecules of which consist of atoms of one element, The ionization process is accompanied by a 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 transferring an electron from a higher energy level to a lower one, and a photon of characteristic radiation is emitted. Thus, 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 the breaking of 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%). By absorbing energy, a water molecule loses an electron, which quickly transfers its energy to surrounding water molecules:

H 2 O = > H 2 O + + e – .

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

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

N + O 2 = > BUT 2 ; BUT 2 + BUT 2 => H 2 O 2 + 2 O.

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 molecules of protein, enzymes, DNA and other cellular structures. Free radical induced chemical reactions develop with great yield and involve in this process many hundreds and thousands of molecules not affected by radiation.

The effect 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 in a time of the order of 10 -16 – 10 -14 s;

2) at the molecular level – physico-chemical changes in macromolecules caused by the direct and radiolytic effects of radiation, leading to disruption of intracellular structures, over a period of about 10 -10 – 10 -6 s;

3) at the biological level – dysfunction of tissues and organs, developing over a period of time from a few seconds to several days or weeks (in case of acute lesions) or over years or decades (long-term effects of radiation).

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

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

In higher organisms, cells are organized into tissues and organs that perform various functions, for example: production and storage of energy, muscle activity for movement, digestion of food and excretion of waste, supply of oxygen, 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.

The random distribution of energy absorption events created by radiation can damage vital parts of the DNA double helix and other cell macromolecules 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 disrupted, the activity of enzyme systems is suppressed, tissue growth slows down and stops, and new ones arise. chemical compounds, not characteristic of the body - toxins. The final unwanted radiation effects are divided into somatic and genetic.

Somatic effects manifest themselves directly in the irradiated person himself or as early detectable effects exposure (acute or chronic radiation sickness and local radiation injuries), or as long-term consequences(reduced life expectancy, the occurrence of tumors or other diseases) that appear 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 descendants. These effects of radiation can be very long-term and extend over several generations of people.

The severity of the harmful effect depends on the specific tissue irradiated and the body's ability to compensate or repair the damage.

The ability to restore cells depends depending on the person's age at the time of irradiation, on gender, health status and genetic predisposition of the body, as well as on the size absorbed dose(radiation energy absorbed per unit mass of biological tissue) and, finally, from type of primary radiation affecting the body.

1.2. Threshold and non-threshold effects during human irradiation

In accordance with modern concepts, set out in publication 60 of the ICRP and forming the basis of the Russian Radiation Safety Standards NRB-99, possible harmful effects of radiation on health are divided into two types: threshold (deterministic) and non-threshold (stochastic) effects.

1.Deterministic (threshold) effects – immediate early, clinically detectable radiation diseases, having dose thresholds below which they do not occur, and above which the severity of the effects depends on the dose. These include acute or chronic radiation sickness, radiation cataracts, reproductive dysfunction, cosmetic damage to the skin, degenerative damage to various tissues, etc.

Acute Radiation sickness occurs after exceeding a certain threshold dose of a single exposure and is characterized by symptoms depending 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 permissible limits. Signs of chronic radiation sickness are changes in blood composition (decreased number of leukocytes, anemia) and a number of symptoms from the nervous system. 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. Increased losses may be compensated by an increase in the replacement rate, but there may also be a temporary and sometimes permanent decrease in the number of cells capable of supporting the function of the organ or tissue.

The resulting cell loss can cause severe impairment 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 noticeably impair tissue or organ function. In addition to cell death, radiation can cause tissue damage in other ways: by affecting numerous tissue functions, including regulation of cellular processes, inflammatory responses, suppression of the immune system, and the 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, deterministic effects of radiation are specific and do not arise under the influence of other physical factors, and the connection between the effect and irradiation is unambiguous (deterministic). Threshold doses for the occurrence of deterministic effects leading to the rapid death of adults are given in Table 1.2. In the case of long-term chronic irradiation, these same effects occur at higher total doses than in the case of single irradiation.

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

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

Table 1.1. Impact of different doses of radiation on adult health

with a single irradiation

Note. Currently, 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 effects resulting in human death

The dependence of survival on 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, this dose (averaged over the entire body) is 3–5 Gy (Gy) for acute exposure (Table 1.2).

In industrial conditions, the occurrence of deterministic effects is possible only in the event of a radiation accident, when the radiation source is in an uncontrollable state. In this case, limiting the exposure of people is carried out by taking urgent measures - intervention. 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 deterministic effects

and criteria for urgent intervention in a radiation accident

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

2. Stochastic, or non-threshold, effects – long-term consequences of radiation 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 cancer and hereditary diseases that arise spontaneously over the years in people for a variety of natural reasons.

The reliability of the connection of a certain part of these effects with radiation was proven by international medical and epidemiological statistics only in the early 1990s. Stochastic effects are usually detected through long time after irradiation and only with long-term observation of large groups of the population 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 radiation varies between men and women and varies with time after exposure (Figure 1.1).

The likelihood of malignant degeneration of a cell is influenced by the size of the radiation dose, while the severity of a certain type of cancer depends only on its type and location. It should be noted that if the irradiated cell does not die, then it has a certain ability to self-repair the damaged DNA code. If this does not happen, then in a healthy body its vital functions are blocked immune system: The degenerated cell is either destroyed or does not reproduce until its natural death. Thus, the likelihood of cancer is low and depends on the “health” of the body’s immune and nervous systems.

Reproduction process cancer cells has a random nature, although due to genetic and physiological characteristics people can vary greatly in their sensitivity to radiation-induced cancer. Some people with rare genetic diseases may be significantly more sensitive than the average person.

With small additions of dose to natural (background) radiation, the probability of causing additional cases of cancer is naturally small, and the expected number of cases that can be attributed to the 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 a spontaneous level of stochastic effects, any practical activity that leads to additional exposure leads to an increase in the likelihood of stochastic effects. The probability of their occurrence is assumed to be directly proportional to the dose, and the severity of the manifestation is assumed to be independent of the radiation dose.

Figure 1.2 illustrates the relationship between radiation exposure and the incidence of cancer in the population. It is characterized by a significant level of spontaneous cancers in the population and a relatively small probability of the occurrence of additional diseases under the influence of radiation. In addition, according to UNSCEAR data, the spontaneous rate of incidence and mortality from cancer varies significantly from country to country and from year to year in one individual country. This means that by analyzing the effects of 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 radiation, however, it is impossible to indicate which disease is a consequence of radiation and which arose spontaneously.

Figure 1.3 shows an estimate of the size of a group of equally exposed adults required 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 a theoretical estimate of the group size required to identify additional stochastic radiation effects with a 90% confidence interval. Above this line is a region in which it is theoretically possible to prove a connection between an increase in the number of stochastic effects in a group and radiation exposure. Below this line, it is theoretically impossible to prove this connection. The dotted line shows that for reliable detection additional effects from uniform irradiation of the body of adults with photons with a dose of 20 mGy, equal to the occupational radiation dose limit, it is necessary to examine at least 1 million people with such a dose.

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

1.3. Basic dosimetric quantities and units of measurement

Activity (A) – a measure of the amount of 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 radionuclide atoms. The value of the total activity characterizes the potential radiation hazard of premises in which work with radioactive substances is carried out.

SI unit of measurement – Bk(becquerel), equal to 1 decay per second ( s –1).

Non-systemic unit – Ki(curie); 1 Ci = 37 GBq = 3.7×10 10 s –1.

Particle flow ( F) – the number of 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 quantity 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, % – coefficient of 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) – the ratio of the number of elementary particles (alpha, beta, photons, neutrons) penetrating into an 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. 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 of ionizing radiation. In nuclear physics, an off-system unit of energy is used - the electronvolt (eV). 1 eV = 1.6020×10 -19 J.

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

SI unit of measurement – C/kg(pendant per kilogram).

Non-systemic unit – R(X-ray); 1 P = 2.58×10 -4 C/g; 1 C/kg = 3872 RUR.

Absorbed dose, or simply dose ( D) – a measure of the physical impact of ionizing radiation on a substance (at the molecular level). Equal to the ratio of the radiation energy absorbed in a substance to form ions to the mass of the irradiated substance.

SI unit of measurement – Gr(gray); 1 Gy = 1 J/kg.

Non-systemic unit – glad(rad – radiation absorbed dose);

1 rad = 0.01 Gy = 10 mGy.

The exposure dose of photon radiation X = 1P 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 of ionization of molecules. For practical radiation safety purposes, we can assume that 1 R corresponds to 1 rad or 10 mGy.

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

N = å D R × W R

Radiation weighting coefficient values W R accepted into 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 of measurement – Sv(sievert); for gamma radiation 1 Sv = 1 Gy.

Non-systemic unit – rem(biological equivalent of rad);

1 rem = 0.01 Sv = 10 mSv.

Relationship with other dose units:

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

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

For neutron radiation, an absorbed dose of 1 rad (10 mGy) will 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 long-term 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 N– the same equivalent dose to all organs and tissues .

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

E = å N T × W T(T = 1 ... 13),

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

Collective dose ( S) – a measure of the potential harm to society from the possible loss of person-years full life population due to the implementation of long-term consequences of radiation. 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 - man-Sv(person-sievert).

To justify the costs of radiation protection in NRB-99, it is accepted that exposure to a collective dose of S = 1 person-Sv leads to potential damage equal to the loss of 1 person-year of 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. Just like flux density, dose rate characterizes the radiation situation (radiation level) at a point in a room or area.

The following abbreviations of the term are often used:

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

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

Natural background is the level of natural gamma radiation, which on average at sea level is due 1/3 to cosmic rays and 2/3 to radiation from natural radionuclides contained in 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 the range = (8–12) µR/h. This corresponds to the flux density j about 10 photons / (cm 2 s), and also:

In MTD units =(8–12) μrad/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, the DER of natural gamma radiation is allowed to exceed 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 around the world, 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 Radiation Safety Standards NRB-99

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

The Standards vary in their capabilities for source control and exposure control. four types of radiation exposure per person :

· from man-made sources under conditions of their normal operation (the source and radiation protection are controlled and managed);

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

· from natural sources radiation (uncontrolled exposure);

· from medical sources for the purpose of diagnosis and treatment of diseases.

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

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

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

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

To ensure radiation safety during normal operation of sources, they are used three main principles of the Republic of Belarus:

· principle of justification – prohibition of all types of activities involving the use of radiation sources, in which the benefit obtained for individuals and society does not exceed the risk possible harm caused by additional exposure;

· rationing principle – not exceeding permissible limits individual doses of radiation to citizens from all sources of radiation;

· optimization principle – maintenance at the lowest possible and achievable level taking into account economic and social factors individual radiation 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 man-made sources);

· Group B personnel(persons who, due to work 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 have no medical contraindications for working with ionizing radiation, who have undergone special training and subsequently undergo an annual medical examination. Group B personnel – persons at least 18 years of age (including students undergoing laboratory practical work with sources). In the “Population” category, as a rule, children aged 0 years and older are distinguished. Many concepts in NRB-99 are standardized, for example, average duration life when considering the risk of non-threshold effects is taken to be 70 years.

· basic dose limits (MD) – such values ​​of the individual annual effective dose, the non-exceeding 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 (LA) – derivatives of the basic dose limits for assessing the radiation situation. At one-factor exposure from external sources is the average annual permissible dose rate in work premises ( DMD );

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

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

4. The effective radiation dose for personnel over a 50-year working period should not exceed 1000 mSv, and for the population over a 70-year life span – 70 mSv.

5. When a person is simultaneously exposed to sources of external and internal radiation (multifactorial irradiation) the main dose limits indicated in Table 1.4 refer to total annual dose, caused by all factors. Therefore, the values ​​of RL (DMD) for each irradiation factor separately should be taken less than in Table 1.5.

6. For women for those under 45 years of age classified as Group A personnel, additional restrictions have been introduced: the equivalent dose to the lower abdominal area should not exceed 1 mSv per month. Under these conditions, the effective dose of radiation to the fetus is 2 months. undiagnosed pregnancy will not exceed 1 mSv. After establishing the fact of pregnancy, the enterprise administration is obliged to transfer the woman to a job that does not involve radiation.

7. Planned increased exposure above the established dose limits (PD = 50 mSv effective dose) is permitted 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. Irradiation in doses above 4 PD (200 mSv) is considered potentially dangerous. Persons exposed to radiation at such doses are allowed to subsequently work with radiation sources only on an individual basis, by decision of the competent medical commission.

Cases unplanned increased exposure people at doses above the PD are subject to investigation.

Table 1.4. Basic Dose Limits

**All PD and PD values ​​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 irradiation

2.1. Preparing for work

Goal of the work

1. Assessing the radiation safety of students and laboratory personnel 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. Reconciliation of readings from various dosimeters with dose rate calculations.

Equipment and materials used

1. A 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. The container is equipped with collimator(an opening channel that allows one to obtain a limited beam of g-radiation).

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

3. Dosimeters with detectors that record gamma radiation.

Main characteristics of an installation with a gamma radiation source

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

To protect against gamma radiation, the GIK-2-9 source is placed in a lead container with a wall thickness of 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, dose rate measurements are made at various distances from the source.

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

· sketch of a protective container with a source (sectional view);

· energy of cobalt gamma radiation photons (Eg = 1.25 MeV);

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

initial activity of the source Ao(Bq) and date of source certification;

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

· value of gamma – cobalt constant - 60 G (nGy × m2/(s × GBk))

2.2. Radiation safety assessment when working with a source

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

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

1. Having walked around the laboratory premises with a dosimeter, perform radiation reconnaissance, i.e. find places with increased level gamma radiation. It is recommended to measure the DER on the surface of all devices marked with radiation hazard signs(containers, safes, source kits on other desktops). Write down the 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 equivalent dose rate r.m at the workplace located in maximum proximity to the source, i.e. With highest level radiation. The collimating channel of the source must be open, i.e. The worst radiation situation has been created. 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 unevenness of irradiation of organs and tissues of the body near the source, i.e. measure the DER T for 13 organs and tissues, and then multiply them by the weighing coefficients of radiosensitivity W T. In our conditions, it is enough to limit ourselves to measurements for four control points of the body: 1 – head, 2 – chest, 3 – gonads, 4 – feet, and accept for them the enlarged weighting coefficients W K (see Table 2.1).

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

= Σ ( K · W K), (2.1)

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

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

Total: =Σ ( K Wk) = ___________ μSv/h

Find the coefficient of irradiation unevenness, equal to the ratio of the effective dose to the readings of one dosimeter:

α = /

and make a conclusion about whether, under these conditions, it is advisable 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 laboratory practical work, determine the maximum possible effective dose of man-made exposure to the student for the current year:

E stud = · 16.

7. For the same reasons, estimate the maximum possible annual dose of group A personnel, taking the standard employee work time to be 1700 hours:

E pers = · 1700.

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

E eats = f · 8760.

Assess the possible spread of natural radiation dose by roughly taking the confidence interval based on the maximum and minimum background values ​​measured in step 2:

Δ = (max – min) 8760,

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

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

r = E stud, pers r E,

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

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

Compare the doses of man-made radiation to employees and students with the expected annual dose from natural radiation and its spread.

2.3. Removing the dependence of dose rate on distance

In this part of the work, it is necessary to remove 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 the collimator open the detector located in the gamma radiation beam “sees” the point source directly and registers its direct radiation. Absorption and scattering in air at short distances can be neglected, so in this case it holds inverse square law: the intensity of radiation in a 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 the collimator closed the detector registers radiation that is significantly attenuated (by a factor of 300 or more) and scattered in the lead shielding. 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 satisfied at large distances from it.

To take measurements the detector of the selected dosimeter is installed on a carriage, which 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, to find the limit where the device does not “go off scale”. In the selected range, take 4–5 dose rate readings at various distances and subtract the background from them . Record the values ​​of distances and dose rates in the observation log (Table 2.2). In the journal, you should convert the dosimeter readings into DER units (μSv/h), if the device is calibrated in other units.

Measurements should be repeated with several instruments with the collimator open and closed. It should be taken into account that due to the different sensitivity of dosimeters, some of them can go “off scale” in an open beam, while others may not show anything in a closed beam. The UIM-2-2 device, calibrated in units of s –1, measures the flux of photons through the detector (F) and is called radiometer. To convert its readings into units of dose rate, you should use the calibration dependencies located on the desktop.

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

Table 2.2. Log of equivalent dose rate measurements

Note: The natural background should be subtracted from the readings marked *.

2.4. Calculation of dose rate based on source activity

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

Table 2.3. Logbook for dose rate calculations

A = 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 certification date, T 1/2 is the half-life (n must be dimensionless); Ao– initial activity of the source according to the passport (data taken from the laboratory poster).

2. Recalculate the rated exposure dose rate on the date of the experiment in the same way 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 various distances from the source located 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 · A/ r 2 , nGr /s, (2.3)

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

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

O(r) = Г A/ r 2 · 3.6 · W R , μSv/h. (2.4)

Write the calculations using formula (2.4) in line number 2 of Table 2.3.

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

4. Make allowances for the attenuation of gamma radiation in the air. Take the thickness of the air layer equal to the distance from the source to the detector, x = r.

The weakening factor of the air layer thickness x B cm is

K = exp (μ V x V) / V ∞,

where μ B is the linear attenuation coefficient of air, depending on the energy of gamma quanta, cm –1; B ∞ is the accumulation factor in infinite geometry, taking into account the contribution of radiation scattered by air (depending on the energy of gamma quanta and on the product μх). Take these values ​​according to tables A.1 and A.2 for the energy of gamma radiation from the source.

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 protection can be considered a barrier). The reduction factor for lead protection with a thickness x P b = 10.5 cm is

К Р b = exp (μ Р b x Р b) / (В Р b d) ,

where μ Р b is the linear attenuation coefficient of lead, taken according to the energy of gamma rays (Table A.1); In P b is the lead accumulation factor for infinite geometry, adopted according to Table A.2, and d is the correction for barrier geometry (depending only on the energy of gamma quanta), adopted according to Table A.3. The DER taking into account the attenuation in lead 2 = 1 / K P 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 versus distance: one graph for the case of an unprotected source - 1 (r), the other for a source placed in a container - 2 (r). For the convenience of comparing dosimeter readings with calculations, experimental points from Table 2.2 should be shown on the graphs.

7. The conclusions for this part of the work include:

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

Think over possible reasons deviations of instrument readings from calculated values;

Assess the absorption capacity of air;

Control questions

1. Effects of ionizing radiation on the human body.

2. Deterministic effects of radiation, development mechanism.

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, area of ​​application.

7. Collective dose and collective damage.

8. Dose rate. Natural radiation background.

9. Radiation safety goals and ways to achieve them.

10. Principles of ensuring radiation safety.

11. Principle of justification.

12. The principle of rationing.

13. Optimization principle.

14 Types of human exposure considered in NRB-99.

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

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

17. Permissible levels for external technogenic irradiation - 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 and activity and distance.

19. Law of attenuation of radiation with distance.

20. 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 irradiation time and permissible dose rate.

24. Allowable time for working with a radiation source (in what cases should it be assessed and how).

Bibliography

2. Federal Law “On Radiation Safety of the Population”. No. 3-FZ dated 01/09/1996.

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

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

5. Kutkov, V.A. Basic provisions and requirements regulatory documents in the practice of ensuring radiation safety of nuclear power plants: textbook / V.A. Kutkov [etc.] – M: Publishing house. OIATE, 2002. – 292 p.

6. Kozlov, V.F.. Handbook of radiation safety / V.F. Kozlov. – M.: Energoatomizdat, 1999. – 520 p.

7. Norms radiation safety NRB-76/87 and Basic sanitary rules for working with radioactive substances and other sources of ionizing radiation OSP-72/87 / Ministry of Health of the USSR. – 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 IN b point isotropic source in barrier geometry ( d = B b/v ∞ )

1. FUNDAMENTALS OF RADIATION SAFETY……………….…………....3

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

1.2. Threshold and non-threshold effects during human irradiation…….…….…5

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

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

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

2.2. Radiation safety assessment when working with a source……….….19

2.3. Removing the dependence of dose rate on distance………………………..21

2.4. Calculation of dose rate based on source activity…………………………..23

Test questions………………………………………………………..25

Bibliography…………………………………………………….…26

Appendix…………………………………………………………………………………..26

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

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