Industrial gamma ray detection health protection standards
Radiological protection standards for industrial gamma
defect detecting
GBZ132-2002
Preface
Chapters 4 to 8 of this standard and Appendix A, Appendix B is mandatory, and the rest is recommended.
According to the "Chinese People's Republic of China *** and the State Law on Prevention and Control of Occupational Diseases" to develop this standard. The original standard GB18465-2001 and this standard is inconsistent with this standard shall prevail.
The preparation of this standard mainly refers to GB14058, DIN54115 part 1 and its annexes and DIN54115 part 5 of the content, and in conjunction with the actual situation in China and the preparation.
This standard appendix A, appendix B is a normative appendix.
This standard by the Chinese people *** and the Ministry of Health and the State put forward and under the control.
This standard drafting unit: Shandong Academy of Medical Sciences, Institute of Radiation Medicine.
The main drafters of this standard: Deng Daping, Hou Jinpeng, Zhu Jianguo, Wen Jihui, Wang Chunliang.
This standard is interpreted by the Ministry of Health of the People's Republic of China.
1 Scope
This standard specifies the gamma-ray flaw detection machine protection performance and its use in the process of radiation protection and related monitoring requirements.
This standard applies to the application of γ-ray flaw detector for non-destructive testing of the internal structure of metal components.
2 Normative references
The provisions of the following documents become the provisions of this standard by reference. All subsequent change orders (excluding errata) or revisions of dated references do not apply to this standard; however, parties to agreements based on this standard are encouraged to investigate the possibility of using the most recent versions of these documents. Where undated references, the latest version is applicable to this standard.
GB4075 Classification of sealed radioactive sources
GB11806 Regulations for the safe transportation of radioactive material
GB/T14058 Gamma-ray flaw detector
3 Terms and definitions
The following terms and definitions apply to this standard.
3.1 mobile defect detecting mobile defect detecting
Outdoor, production workshop or installation site with portable or mobile γ-ray flaw detector for the detection of the work process.
3.2 Stationary defect detecting stationary defect detecting
The process of γ-ray flaw detection in a dedicated γ-ray flaw detection room with a fixed installation or a limited mobility flaw detector.
3.3 gamma defect detecting room gamma defect detecting room
Placement of gamma ray detector and the object to be examined to carry out gamma ray detection and have a certain shielding effect of the special irradiation room.
4 Gamma ray detector radiation protection performance requirements
4.1 Source container should be in line with GB/T14058 in Article 5.3 of the test requirements, the air around the specific kinetic energy rate does not exceed the value in Table 1.
Table 1 source container around the air than the release of kinetic energy rate control value (mGy-h-1)
The type of flaw detector
from the outer surface of the container
Container outer surface
50mm
1m
Portable
2
0.5
0.02 <
Mobile
2
1
0.05
Fixed
2
1
0.10
4.2 When depleted uranium is used as the shielding material of the source container, the protection against beta rays shall comply with the requirements of clause 5.3.1 of GB/T14058.
4.3 The source container of each γ-ray flaw detector and its sealed source must be marked in accordance with the requirements of Article 8.1.1 and 8.1.2 of GB/T14058.
4.4 γ-ray flaw detector safety locks, interlocking devices, the source of the position indicator, the system failure of the safety device, to prevent unauthorized operation of the device and other safety devices in accordance with the performance of GB/T14085 in Article 5.4 requirements.
4.5 The safety of the source should be consistent with GB/T14085 in Article 5.5 requirements.
4.6 According to the different needs, the length of the radioactive source transmission device should be as short as possible, and the radioactive source must be able to return to the source container and enter the closed state immediately after each photograph.
4.7 Product specifications should indicate the model, specifications and the main technical indicators and equipment maintenance, storage, transportation methods, should also include: the type of radioactive sources used, characteristics, source containers, leakage dose level on the outside surface, safety measures, automatic shutdown function and common accidents and other content.
5 stationary radiation protection requirements
5.1 γ-ray detection room building (including radiation protection walls, doors, windows, radiation protection labyrinth) should be given full consideration to direct radiation, scattering and shielding materials and structures, and other factors and in accordance with the requirements of this standard Appendix A (normative appendix) to determine the thickness of protection.
5.2 Radiation protection wall 5cm outside the dose rate should be less than 2.5μGy-h-1.
5.3 Radiation protection doors must be fixed at the entrance to the radioactive danger sign, exposure during the period of eye-catching "no entry" warning signs; the entrance to the detection room and the object to be probed out of the entrance to the detection room must be set up acoustic and visual alarm device, the device in the γ-ray detector should be automatically turned on, and can be used to detect the presence of the γ-ray detector. The device should be automatically connected, and can automatically return the radioactive source to the source container when someone passes through; the protective performance of the radiation protection door should be the same as the same side of the wall, and the dose rate of the outer 5cm should be less than 2.5μGy-h-1, and the installation of the door-machine interlocking device and the work of the indicator light; the machine room to install a fixed dosimetry at the appropriate location.
6 Requirements for radiation protection of mobile flaw detection
6.1 Before carrying out flaw detection operation, the workplace must be divided into control area and supervision area.
6.2 The specific kinetic energy rate of the air outside the boundary of the control area should be less than 40μGy-h-1. A clearly visible warning sign of "No Entry to the Radioactive Workplace" must be hung at the boundary. Unauthorized persons shall not be permitted to enter the area, which may be managed manually by means of ropes, chains and similar methods or by supervisory personnel. The method of calculating the extent of the control zone is shown in Appendix B (normative appendix).
6.3 The supervisory area is located outside the control area and is accessible to personnel, trainers or visitors. The boundary dose should be no greater than 2.5 μGy-h-1, and the boundary should be marked with a "Caution, Ionizing Radiation" warning sign, and the public should not be allowed to enter the area.
6.4 The distance between the gamma-ray detector and the object to be examined, the irradiation direction, time and shielding conditions must be taken into account to ensure that the irradiated dose to the operator is lower than the annual dose limit value and should be as low as reasonably achievable.
7 safety requirements for radioactive sources
7.1 Sealed sources selected in accordance with the level of GB4075 selected, unprotected sources for the 43515 level, the source in the device for the 43313 level.
7.2 Replacement of radioactive sources should be approved by the local radiation health protection department and carried out under the supervision of protection professionals, in a fully shielded device, using a remote gripper and support device.
The transfer of a sealed source from a transport container into a source container or from a source container into a transport container must be carried out by means of auxiliary equipment that facilitates the changeover operation and by means of devices with adequate shielding properties. The equivalent dose received by the operator during a change should not exceed 0.5mSv.
7.3 The replacement of the source tray should be approved by the competent authority of the user unit and the local radiation health supervision department. If the loading and unloading of source containers with radioactive sources and source trays is performed by means of a pusher, a suitable change container with adequate shielding shall be utilized.
7.4 Discarded radioactive sources are handled or disposed of in accordance with national regulations and detailed records are kept on file.
7.5 Transportation of radioactive sources in accordance with the relevant provisions of GB11806.
7.6 Source containers or radioactive sources should be stored in a dedicated radioactive source library.
7.7 In the local radiation health protection department under the guidance of the unit should develop a suitable emergency plan and make appropriate emergency preparedness, the plan includes: work procedures, organizational structure, personnel training, emergency planning exercises, emergency facilities.
7.8 The operation site must be equipped with appropriate emergency protective equipment, such as: sufficient shielding thickness of the protective shelter, tunnel shielding block, handle length of not shorter than 1.5 meters of clamps, the appropriate length of wire, pool, sandbags and so on.
8 Radiation Protection Monitoring
8.1 Personal Dose Monitoring for Operators
8.1.1 Gamma-ray flaw detection operators must be routine personal dose monitoring, and the establishment of personal dose files and health management files, the personal annual dose limits are as follows:
a) 5 consecutive years, the annual average effective dose of 20mSv;
a) an average annual effective dose of 20 mSv over 5 consecutive years;
b) an effective dose of 50 mSv in any single year;
c) an equivalent dose to the crystalline lens of the eye of 150 mSv in a year; and
d) an equivalent dose to the limbs (hands and feet) or the skin of 500 mSv in a year.
8.1.2 Accidental dose monitoring for workers should also be carried out, and a detailed record should be kept.
8.2 γ-ray flaw detector protection performance monitoring
8.2.1 Production of γ-ray flaw detector, should be in accordance with the requirements of GB/T14058 type test and factory inspection.
8.2.2 by the use of the unit where the radiation health technical service organizations in accordance with this standard chapter IV of the radiation protection performance requirements of the γ-ray flaw detector acceptance test, which is required by Article 4.1 of this standard shielding effect of the test according to GB/T14058 in Article 6.1, qualified before use.
8.2.3 The use of units should always be on the performance of the safety device for testing, radiation health technical service organizations once a year.
8.2.4 After the detector is moved, the part-time protection personnel must use the appropriate instruments to test the performance of the safety device.
8.2.5 Leak testing of sealed radioactive sources is performed annually by the protection authority.
8.3 Protection monitoring of the workplace
8.3.1 Protection monitoring of the stationary detection workplace
8.3.1.1 The acceptance test must be carried out before the commissioning of the detection room, and it can be used only after passing the test.
8.3.1.2 Before each day's work, the flaw detection workers should check the performance of safety devices, interlocking devices and the status of warning signals and symbols. Check whether there are people staying in the flaw detection room.
8.3.1.3 At the end of each flaw detection operation, the operator shall verify that the source is returned to a safe location with a reliable radiation instrument. The entry and exit of source containers should be monitored and documented.
8.3.1.4 The radiation level of the operation place and the adjacent area of the flaw detection room should be measured once a year by the radiation hygiene technical service organization of the place where the user is located, and the evaluation or improvement suggestions should be made according to the results of the measurements. When the activity of the radioactive source increases, the above radiation levels should be re-measured, and based on the results of the measurement to make appropriate improvements.
8.3.2 Radiation protection monitoring of mobile flaw detection operations
8.3.2.1 Prior to each flaw detection operation, the flaw detection machine shall be inspected in accordance with Section 8.3.1.2 of this Standard and the control area shall be inspected to ensure that no personnel are present in the control area prior to exposure of the source.
8.3.2.2 When the workplace is activated, the radiation level shall be measured around the boundary of the control area and adjusted so that it does not exceed 40 μGy-h-1.
8.3.2.3 The establishment of the operation site radiation patrol system, regular observation of the location and status of radioactive sources.
8.3.2.4 After the end of the flaw detection operations should be carried out in accordance with Article 8.3.1.3 of this standard.
Appendix A
(Normative Appendix)
Determination of shielding
A.1 Principles
A.1.1 The direction of the useful wire bundle shall be considered in determining the shielding. If there is no limitation on the direction of the useful harness, the protective layer for all directions shall be determined according to A.2. If the useful harness is only in a limited direction, then, except for this limited direction, the protective layer shall be determined in accordance with section A.2, and the leakage radiation protective layer in all other directions shall be determined in accordance with section A.3.
A.1.2 The total attenuation of multiple layers of shielding composed of different shielding materials is equal to the product of the attenuation of the individual layers.
A.2 Layers of protection against useful radiation
A.2.1 Calculation of the required attenuation of useful radiation FN in accordance with equation (1),
-............... ........................(A.1)
Formula:KN is the measured or calculated attenuation of useful radiation according to Section A.2.2 calculates the specific kinetic energy rate (mGy/h) in the useful radiation beam at distance a0(m) from the source, a is the distance (m) from a point of the source, and KG is the maximum permissible specific kinetic energy rate (mGy/h) at distance a from the source.
A.2.2 At a distance of a0, the maximum specific kinetic energy rate, KN, at that point is calculated from the expected maximum activity of the source, A(GBq), and the specific kinetic energy constant, TK (see Table A.1), in accordance with Equation (2).
............ (A.2)
Table A.1 Specific kinetic energy constants гK, (mGy-m2)/(h-GBq)
Radiation source
60Co
192Ir
гK
0.35
0.13
A.2.3 Thickness of the protective layer against the beam of useful radiation. The thickness of the protective layer against useful radiation beams can be found in Figures A.1 and A.2. The thickness of the shielding layer in cm is obtained by dividing the mass thickness given in Figures A.1 and A.2 by the density of the shielding material (g/cm3) (see A.2.4 for details).
A.2.4 Formula calculation of the protective layer
The thickness of the protective layer, d(cm), can also be calculated using the value of the linear attenuation coefficient, μ, in Table A.2, according to Equation (3), which is used strictly for the linear range of the curves, FN>10, in Figures A.1 and A.2.
.................. (A.3)
A.2.5 A description of all shielding walls against useful radiation beams, including wall thickness, name of shielding material and thickness, shall be shown on the structural drawings of radiation protection.
A.3 Layers of protection against leakage of radiation
Layers of protection against leakage of radiation from source containers or shields shall be calculated in accordance with formula (4) for the required attenuation FD:
..................... (A.4)
Formula: KD for the useful beam outside the distance from the radioactive source for a0 specific kinetic energy rate (mGy/h).
a0 is the distance from the source to the protection site (m).
KG is the maximum permissible specific kinetic energy rate (mGy/h) at a distance a(m) from the source.
Table A.2 Linear attenuation coefficients
Materials
Linear attenuation coefficient μ(cm-1)
60Co
192Ir
Lead
0.565
1.484
Lead glass
0.231
Iron
0.3095
0.535
General concrete
0.0995
0.137
Barite concrete
0.1385
0.19
Figure A.1 Useful beam attenuation of FN for 60Co, scattered rays attenuation degree of Fs, leakage radiation attenuation degree of FD when the mass thickness of different materials
Figure A.2 192Ir useful beam attenuation degree of FN, scattered rays attenuation degree of Fs, leakage radiation attenuation degree of FD when the mass thickness of different materials
Appendix B
(normative appendix)
Control Zone Determination
B.1 According to the radiation source of gamma rays in all directions when the radiation of different situations, should determine three different types of control zone distance, as shown in Figure B.1.
Figure B.1 control area of the application of shielding (not to scale)
a Ⅰ: radiation without any attenuation of the required control area distance;
a Ⅱ: the direction of the useful beam, by the object of the required control area distance after shielding;
a Ⅲ: the direction of the useful beam outside the direction of the required control area distance by the source container or other shielding
A Ⅲ: the source container or other shielding required control
area distance.
B.
B.2 For mobile flaw detection, the equivalent dose rate at the boundary of the control zone is 40 μSv/h, and the size of each type of distance from the control zone can be evaluated as follows:
aⅠ: the distance from the control zone taken from Fig. B.2 (m)
aⅡand aⅢ: the distance from the control zone taken from Fig. B.2 (m), aⅠ (m), multiplied by a factor corresponding to the different number of half-reduced layers in Table B.2. The product of (can be based on the thickness of the shielding material, divided by the corresponding nuclide and shielding materials in Table B.1 half-reduced layer thickness, to find its half-attenuation layer, and then from Table B.2 to find the corresponding factor).
Table B.1 Approximate values of half-attenuation layer thicknesses for different materials
Shielding materials
Half-attenuation layer thicknesses (HVL) for different radioactive sources (mm)
60Co
192Ir
169Yb
170Tm
Aluminum
70
50
27
20
Concrete
70
50
27
Steel
24
14
9
5
lead
13
3
< p>0.80.6
Tungsten
10
2.5
0.09
Uranium
6
2.3
0.035
Table B.2 Factors for aII and aIII at attenuating radiations when used for control area determination
Number of half-attenuated layers
Factor
0.5
0.9
1
0.7
1.5
0.6
2
0.5
3
0.4
4
0.3
5
0.2
8
0.1
10
0.05
12
0.01
B.3 Examples are as follows:
192Ir,Activity1.85×1012Bq,The test object is a structural steel with a thickness of 28mm (2HV). 28mm (2HVL), radioactive source shielding (irradiation container wall) for tungsten, 25mm thick (10HVL)
a Ⅰ: Fig. B.2 of the control area a Ⅰ = 78m
a Ⅱ: Fig. B.2 of the value of the control area a Ⅰ multiplied by a factor of Table B.2
a Ⅱ = 0.5 × a Ⅰ = 0.5 × 78 = 39m
a Ⅲ : Control area value aⅠ of Fig. B.2 multiplied by the factor of Table B.2
aⅢ=0.05×aⅠ=0.05×78=3.9m
Figure B.2 Distance of the control area when applying gamma sources of different activities without any attenuation of the radiation aⅠ