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 quanta. 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
Option "a".
The effect of radiation on the human body is characterized by the absorbed dose of radiation
where I γ is the full gamma constant of a given radioactive isotope, p cm 2 / mCi h.
C – source activity, mCi, t – exposure time, h;
R is the distance from the source to the irradiated object, cm. The transition from activity (microcuries) to gamma equivalents (in milligram equivalents of radium G) and vice versa is made according to the relationship with I γ = G 8.25, where 8.25 – ionization constant of radium.
t = 41 – number of hours of work per week.
When determining the thickness of the screen, we proceed from the need to minimize the intensity of the radiation flux. For persons of category A (personnel - professional workers directly working with sources of ionizing radiation), the maximum permissible dose (MAD), determined by the "Radiation Safety Standards NRB - 76 and the basic rules for working with radioactive substances and other sources of ionizing radiation OSP - 72/80 is equal to 100 mrem/week
1 rem is a unit of dose of any type of ionizing radiation in the biological tissue of the body, which causes the same biological effect as a dose of 1 rad of x-ray or gamma radiation.
1 rad is an off-system unit of absorbed dose of any ionizing radiation: 1 rad = 0.01 J/kg.
For gamma radiation, the rem is numerically equal to 1 roentgen.
Therefore, traffic allowance = 100 mr/week. The calculated radiation intensity is 54 r/week, i.e. exceeds the permissible limit of 54 · 0.1 = 540 times. This means that the screen must provide attenuation of the radiation intensity by K = 540 times. That's why:
Option "B".
Estimated radiation dose 
r/h,
where M – γ isotope equivalent in mg – Ra equivalent; 8.4 – γ – constant Ra with a platinum filter 0.5 mm thick, p cm 2 / mCi h.
R – distance from the source to the workplace, cm.
The maximum permissible absorbed dose rate for an operator of category "A" is P 0 = 0.1 r/week = 100 / t, mr/h.
where: t – working time in weeks, with a 6-hour working day t = 30 hours.
Required attenuation ratio
Required attenuation ratio taking into account the safety factor
where n is safety factor ≥2.
The thickness of the screen to attenuate the radiation flux by 3.9 times is determined by the formula:
where is the linear attenuation coefficient of radiation by the screen material.
To attenuate radiation with a high atomic number to a high density, the following are suitable for their protective properties: a) stainless steel; b) cast iron; c) concrete; d) tungsten: e) lead.
Let us take the isotope energy for p-radiation to be 3 M3B. Using reference data for radiation energy P = 3 MzV, we determine the linear attenuation coefficients (Table 8.c181):
for iron: f = 0.259 cm –1;
for concrete: b = 0.0853 cm –1;
for tungsten: in = 0.786 cm –1;
for lead: c = 0.48 cm –1.
The thicknesses of the screens, calculated for 3.9 times attenuation of radiation with a safety factor of 2, from the materials considered will be equal to:
a) iron: 
b) concrete: 
c) tungsten: 
d) lead: 
Thus, for a stationary screen, the most practical and cheapest would be a concrete screen with a thickness of at least 24 cm; for mobile screens, lead with a thickness of at least 4.3 cm, iron with a thickness of at least 8.0 cm, or tungsten with a thickness of at least 2.65 cm can be used; for a collapsible metal screen, you can use metal arrow-shaped blocks (cast iron bricks) with a wall thickness of at least 8 cm.
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. In addition, 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 quanta. 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 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 required thickness of the protective layer of a given material x (cm) using universal tables, you 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 value 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. Therefore, 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
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. Valid SanPiN 2.6.1.2523-09 (NRB-99/2009). - 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 of industrial premises with 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” are allowed to work in the Republic of Kazakhstan. 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 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. Valid SP 2.2.1.1312-03, 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 device of the RK 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 the RK 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 pool 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 personal protective equipment to prevent the penetration of these substances and compounds onto the 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 radiation safety service of the organization must develop detailed instructions in case of emergency situations, taking into account the specifics of the design of the Republic of Kazakhstan 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.
