GBZ 129-2002 Occupational internal exposure personal monitoring specification
Some standard content:
Ics13.100
National occupational health standard of the People's Republic of China GBZ129-2002
Specifications of individual monitoring for occupational internal exposure
Specifications of individual monitoring for occupational internal exposure Issued by
Ministry of Health of the People's Republic of China
Terms and definitions
Monitoring methods and their selection
Routine individual monitoring
Special monitoring and task-related monitoring
Interpretation of measurement results
Uncertainty and quality assurance of internal exposure monitoring Appendix A (Normative Appendix) m(t) and m(T/2) of commonly used radionuclides under inhalation conditions
This standard is formulated in accordance with the Occupational Disease Prevention and Control Law of the People's Republic of China Chapter 3 of this standard is mandatory content, and the rest is recommended content. This standard is mainly based on the publications No. 60, 75 and 78 of the International Commission on Radiological Protection (ICRP) and the Safety Series No. RS-G-1.2 of the International Atomic Energy Agency (IAEA). Appendix A of this standard is a normative appendix.
This standard is proposed and managed by the Ministry of Health.
The drafting unit of this standard: China Institute of Radiation Protection. The main drafter of this standard: Zhou Yongzeng.
This standard is interpreted by the Ministry of Health.
National Occupational Health Standard of the People's Republic of China Specifications for Individual Monitoring of Occupational Internal Exposure
Specifications of individual monitoringforoccupational internal exposure—Scope
GBZ129-2002
This standard specifies the basic requirements for the principles, methods, plans and interpretation of measurement results of individual monitoring of internal exposure of radiation workers.
This standard only applies to individual monitoring of internal exposure of occupational exposure. 2 Terms and Definitions
The following terms and definitions apply to this standard. 2.1 Individual monitoring of internal exposure Monitoring of the types and activities of radionuclides in the body or excreta, as well as monitoring of the types and activities of inhaled radionuclides using personal air samplers or respirators worn by workers (hereinafter referred to as personal monitoring unless otherwise specified).
2.2 Intake intake
The amount of radionuclides that enter the body through inhalation or ingestion, or through intact skin or wounds. 2.3 Type F material type F material
Material that enters body fluids from the respiratory tract at a fast absorption rate, all of which is absorbed into body fluids with a biological half-life of 10 minutes.
2.4 Type M material type M material
Material that enters body fluids from the respiratory tract at a medium absorption rate, 10% of which is absorbed with a biological half-life of 10 minutes, and 90% of which is absorbed with a biological half-life of 140 days. 2.5 S-type material type S material
A relatively insoluble substance that enters the body fluid from the respiratory tract at a slow absorption rate, 0.1% of which is absorbed with a biological half-life of 10 minutes, and 99.9% of which is absorbed with a biological half-life of 7000 days. Approved by the Ministry of Health of the People's Republic of China on XXXX-XX-XX and implemented on 4
XXXX-XX-XX
2.6 Personal air sampler personal air sampler (PAS) A portable device specially designed to measure the time-integrated activity concentration of radioactive aerosols or gases in the breathing zone of workers to estimate the intake of the workers. 2.7 Fixed air sampler static air sampler (SAS) A device used to monitor workplace conditions and can provide useful information on the composition and particle size of radionuclides. 2.8 Investigation level investigation level (IL) refers to a specified value of quantities such as effective dose, intake or pollution level per unit area or unit volume, and an investigation should be conducted when this value is reached or exceeded.
2.9 Recording level (RL) A level of dose, exposure or intake specified by the regulatory authority. When the dose, exposure or intake received by the workers reaches or exceeds this level, it should be recorded in their personal exposure records. 2.10 Controlled area controlled area
An area divided within the radiation workplace where special protective measures and safety measures are required or may be required to:
a) Control normal exposure or prevent the spread of contamination under normal working conditions; b) Prevent potential exposure or limit its extent. 2.11 Supervised area supervised area
Any area that is not identified as a controlled area and does not usually require special protective measures and safety measures but whose occupational exposure conditions are constantly checked.
2.12 Routine monitoring routine monitoring Monitoring conducted at predetermined time intervals at a predetermined location to determine whether the working conditions are suitable for continued operation.
2.13 Task-related monitoring Task-related monitoring is used for specific operations and is intended to provide data for current decisions of operational management. It can also be used to support protection optimization.
2.14 Special monitoring special monitoring Monitoring performed over a limited period to clarify a specific problem. 3 General
3.1 Monitoring The main purpose of personal monitoring of internal exposure is to estimate the committed effective dose and, when necessary, the committed equivalent dose to severely exposed tissues to verify compliance with regulatory requirements;
b) assist in the design and operational control of facilities; and c) provide valuable data for initiating and supporting any appropriate health surveillance and treatment in the event of accidental exposure. 3.2 Monitoring principles
For workers who work in controlled areas and may have significant intake of radionuclides, routine personal monitoring should be carried out: if possible, all workers exposed to occupational exposure should be subject to personal monitoring, but if experience shows that the committed effective dose generated by the annual intake of radionuclides is unlikely to exceed 1 mSv, personal monitoring is generally not required, but workplace monitoring should be carried out.
3.3 Monitoring methods
The personal monitoring methods used to estimate the intake of radionuclides include: a) direct measurement of radionuclides in the whole body or organs; b) analysis of excreta or other biological samples; c) analysis of air sampling.
Each measurement method should be able to identify and quantify radionuclides, and its measurement results can be explained by intake or committed effective dose.
3.4 Monitoring types
According to the purpose of monitoring, personal monitoring can be divided into routine monitoring, special monitoring and task-related monitoring. Wound monitoring and post-medical intervention monitoring are both special monitoring.
4 Monitoring methods and their selection
4.1 Direct measurement of radionuclides in the whole body or organs 4.1.1 Direct measurement techniques for the content of radioactive substances in the whole body or organs can be used for radionuclides that emit characteristic X-rays, gamma rays, positrons and high-energy beta particles, and can also be used for certain alpha radiators that emit characteristic X-rays. 4.1.2 The equipment used to directly measure the content of radionuclides in the whole body or organs consists of one or more high-efficiency detectors installed in a low-background environment. The geometric position of the detector should be consistent with the measurement purpose. For fission products and activation products that emit gamma rays, such as 131I, 137Cs and 6°Co, simpler detectors that can be used in the workplace can be used for monitoring. For a few radionuclides such as isotopes of cyclopentane, high-sensitivity detection techniques are required. 4.1.3 Radioactive substances in wounds that can emit high-energy gamma rays can usually be detected by beta-y detectors. When the contaminant 6
is certain alpha radiators that can emit characteristic X-rays, it can be detected by X-ray detectors. When the wound is contaminated with multiple radionuclides, a detector with energy discrimination should be used. The wound detector should be equipped with a good collimator to locate the radioactive contaminants.
4.1.4 Decontamination of the human body surface should be performed before direct measurement 4.2 Analysis of excreta and other biological samples
4.2.1 For radionuclides that do not emit gamma rays or emit only low-energy photons, excreta monitoring may be the only appropriate monitoring technique. For radiators that emit high-energy beta and gamma rays, excreta analysis is also a commonly used monitoring technique. Although in some cases, such as when elements are mainly excreted through feces or to evaluate the clearance of inhaled S-type substances from the lungs, fecal monitoring programs generally only include urine analysis. 4.2.2 Analysis of other biological samples is for special investigations. For example, nasal mucus or nasal swab samples can be analyzed as a routine screening technique; blood samples can be analyzed as appropriate when high-level contamination is suspected: in the case of internal contamination of 226Ra and 228Th, exhaled breath activity measurement is a useful monitoring technique. In the case of wounds contaminated with extremely toxic radionuclides (such as transuranic elements), samples of excised tissue should be prepared and/or measured in their original form. 4.2.3 The following should be noted when collecting, storing, handling and analyzing urine samples: a) The collection, storage, handling and analysis of urine samples should avoid external contamination, cross contamination and loss of the nuclides to be tested; b) For most routine analyses, 24-hour urine should be collected. In routine monitoring, if 24-hour urine cannot be collected, the urine volume should be corrected to 24-hour urine using creatinine or other amounts; xenon is an exception. Generally, only a small amount of urine can be collected to infer the body fluid concentration and intake from the measured urine xenon concentration.
c) The volume required for analysis is related to the sensitivity of the analytical technique. For some radionuclides, it is necessary to analyze urine samples accumulated over several days to achieve the required sensitivity; d) Sample processing and analysis should be carried out in accordance with relevant standard methods. e) In some cases (such as special monitoring), in order to reduce the impact of the daily excretion fluctuations of the nuclides excreted through urine on the monitoring results, urine samples from three consecutive days should be analyzed separately, or mixed samples from three consecutive days should be analyzed, and the average value should be used as the daily excretion of the middle day. 4.2.4 Due to the large fluctuations in daily fecal excretion of radionuclides, the interpretation of routine fecal sample monitoring data contains greater uncertainty. Therefore, fecal samples should be collected for several consecutive days. Fecal sample monitoring is often used in special investigations, especially after known or suspected inhalation of M or S substances. In these cases, the measurement of daily fecal excretion is very helpful for evaluating clearance from the lungs and estimating intake. The precautions in 4.2.3 also apply to fecal samples. 4.2.5 Gamma emitters in biological samples can be directly measured using scintillation detectors or semiconductor detectors. For alpha and beta emitters, chemical separation is required and then measurement is carried out using appropriate measurement techniques. Measurement of total alpha or total beta activity in a sample is sometimes useful as a simple screening technique, but it cannot be used to quantitatively estimate intake or committed effective dose unless the composition of the radionuclides is known.
4.3 Air sampling and analysis
4.3.1 The estimation of intake based on the measurement results of air samples carries a large uncertainty. For radionuclides that do not emit strong penetrating radiation and have very low concentrations in excreta, such as steel series elements, the measurement results of air samples can be used to estimate the intake. 4.3.2 The sampling head of PAS should be within the breathing zone, and the sampling rate should preferably represent the typical breathing rate of the staff (~1.2mh). The radioactivity on the filter membrane can be measured by non-destructive technology at the end of the sampling cycle to promptly detect abnormal high levels of exposure. Then the filter membrane is retained, and the filter membranes accumulated for a long time are combined together and measured using radiochemical separation and extraction methods and high-sensitivity measurement techniques. 4.3.3 The requirements for PAS are as follows:
a. Sufficient radioactive materials should be collected. The amount of collection mainly depends on the requirements for the minimum committed effective dose that can be monitored by PAS. For routine monitoring, it is generally required to monitor the committed effective dose generated by the annual intake to exceed 1/10 of the annual dose limit:
b. The sampler should extract a sufficient volume of air to give a numerical value that meets statistical requirements for the air activity concentration in the worker's breathing zone;
c. The particle collection characteristics of the sampler should be known. 4.3.4 PAS does not provide information on particle size, and the particle size has a significant impact on the estimation of particle deposition in the respiratory tract and its dose, so the distribution of inhaled particle size should be determined by actual measurement or a realistic assumption should be made about the particle size distribution. In the absence of special information on particle size, the activity median aerodynamic diameter (AMAD) can be assumed to be 5μm. 4.3.5 For compounds that are easily diffused in the air, such as radioactive gases and vapors (such as 14CO2 and ammonium water), SAS can give a more reasonable estimate of their inhalation amount. For other substances, such as resuspended particles, the error given may be one order of magnitude or more.
4.3.6 By comparing the PAS and SAS measurement results, the ratio between the two is determined, and the ratio can be used to interpret the SAS measurement results. When using the SAS measurement results to estimate personal doses, it is required to carefully evaluate the irradiation conditions and work practices. 4.4 Principles for the selection of monitoring methods
4.4.1 When selecting a monitoring method, the following factors should be considered: a) The radiation characteristics of the radioactive nuclides;
b) The biodynamic behavior of the contaminants:
c) Considering the retention characteristics of the contaminants in the body after biological clearance and radioactive decay: d) The required measurement frequency;
e) The sensitivity and convenience of the measurement equipment under consideration and whether such equipment is available. 4.4.2 For routine monitoring, if the sensitivity can be met, generally only one measurement technology is used. For the box, only urine xenon analysis is used. For other nuclides, such as isotopes of rings, different measurement methods should be used in combination because of certain difficulties in measurement and data interpretation. Special monitoring often uses two or more monitoring methods. 4.4.3 Considering the accuracy of data interpretation, the three monitoring methods described in Chapter 4 are generally selected in the following order: direct measurement of radionuclides in the whole body or organs, analysis of excreta and other biological samples, and analysis of air sampling. 5 Routine personal monitoring 5.1 Application of routine monitoring According to Article 3.2, routine personal monitoring should generally be carried out in the following situations: a) Operation of large quantities of gaseous and volatile substances, such as xenon and its compounds produced in large-scale production processes: b) Processing of cyclopentane and other transuranic elements: c) Mining, beneficiation and processing of uranium and the application of cyclopentane and its compounds d) Mining, beneficiation and processing of high-grade uranium ores: e) Processing of natural uranium and low-enriched uranium and production of reactor fuel: f) Large-scale production of radioactive isotopes: g) Working in uranium mines and other workplaces where hydrogen levels exceed action levels: h) Handling large quantities of 131I-labeled radiopharmaceuticals: i) Reactor maintenance that can cause irradiation of fission and activation products: j) For new operations.
5.2 Frequency of routine monitoring
5.2.1 The frequency of routine monitoring is related to the retention and discharge of radionuclides, the sensitivity of the measurement technology, the type of radiation, and the acceptable error in the estimation of intake and committed equivalent dose. 5.2.2 When determining the monitoring frequency, the intake underestimated by more than 3 times due to the assumption that the intake occurred on the middle day of each monitoring period due to the unknown intake time. 5.2.3 In general, the selection of the monitoring period should not result in the omission of intakes corresponding to more than 5% of the annual dose limit. 5.2.4 In principle, sensitive measurement methods should be used as much as possible, but before the measurement method is selected, the cost of using the most sensitive detection technology and the shortest possible sampling period should be weighed against the radiation hazards caused by the underestimated or missed dose due to the use of less sensitive detection technology or longer monitoring period. 6 Special monitoring and mission-related monitoring
6.1 Since special monitoring and mission-related monitoring are related to special events that actually occurred or are suspected to have occurred, the intake time is known and it is also possible to obtain information on the physicochemical state of the contaminants. The relevant provisions for interpreting routine monitoring results do not apply to special monitoring and mission-related monitoring. 6.2 Special monitoring is required when ingestion is known or suspected, or after an accident or abnormal event. Special monitoring is also often carried out because the results of routine excretion measurements exceed the derived investigation level and abnormalities are found in samples collected temporarily such as nasal and nasal swabs and other monitoring results.
6.3 Wound monitoring is a special monitoring. In this case, the amount of radioactive material in the wound site should be determined. If resection 9
surgery has been performed, the radioactive material in the resected tissue and remaining in the wound site should be measured. Then direct urine and fecal excretion monitoring should be performed as needed.
6.4 Monitoring after medical intervention is a special monitoring. If absorption-blocking or excretion-promoting drugs are used, the relevant data recommended in the annex to ICRP Publication 78 cannot be directly used to deduce the committed effective dose. If such treatment is carried out after an accidental ingestion, a special monitoring plan should be developed to track the distribution, retention and excretion of the contaminant in the body of the accidental ingestor, and based on these data, a special estimate of the committed effective dose for the ingestor should be made. 6.5 When the committed effective dose generated by the intake of radionuclides approaches or exceeds the annual dose limit, data on the exposed individuals and pollutants are generally required, including the physical and chemical state of the radionuclides, particle size, retention characteristics of the nuclides in the exposed individuals, nasal and skin contamination levels, air activity concentrations and surface contamination levels, etc. These data are then analyzed and used to give a reasonable intake estimate.
7 Interpretation of measurement results
7.1 Routine monitoring
7.1.1 For routine monitoring, assuming that the intake occurs at the midpoint of the monitoring period T (days), the intake I should be calculated using the measurement value M not obtained during the monitoring period according to the following formula:
I=M/m(T/2)
Where:
I is the intake of radionuclides in becquerel (Bq):.)
M is the content of nuclides in the body or organs measured t days after intake (Bq), or the daily excretion (Bqd): m(T/2) is the expected value of the content of nuclides in the body or organs (Bq), or the daily excretion (Bqd\) T/2 days after the intake of a unit activity. The activity here can also be expressed as a fraction of the intake (except for water). For the values of m(T/2) of some commonly used radionuclides, see Appendix A (Normative Appendix). 7.1.2 Intakes generated in previous monitoring cycles may affect the measurement results of the current monitoring cycle. If more than about 10% of the current measurement value comes from intakes in previous monitoring cycles and the intake and dose have been estimated, the measurement results of the current monitoring cycle should be corrected. For a series of measurements in a routine monitoring plan, the following steps can be followed: a) determine the intake value of the first monitoring cycle; b) estimate the contribution of this intake to the measurement results of subsequent monitoring cycles; c) subtract this contribution from the data of subsequent monitoring cycles; d) repeat a) to c) for the next monitoring cycle. 7.1.3 In a routine monitoring plan, if the monitoring results exceed the pre-determined investigation level, further investigation should be carried out. The nature of the investigation will depend on the specific circumstances and the extent to which the monitoring results exceed the investigation level. In the investigation, the following points should be considered: a) Repeat measurements to confirm or improve the initial assessment; b) Use additional monitoring techniques:
c) Evaluate working conditions and exposure situations;
d) If default parameter values were used in the initial assessment, if necessary, investigate the actual particle size and chemical form of the contaminants and select more appropriate values: e) In the case of large intakes, remove the contaminated person from radioactive work and monitor the retention and excretion characteristics of the contaminants in the body of the intake person to improve the dose assessment. 7.2 Special monitoring and task-related monitoring
In this case, the time of intake is known. If only one measurement is made, the intake I can be calculated by the following formula: I = M/m(t)
Where:
I is the intake of radionuclides, in becquerels (Bq): ·(2)
M is the content of the nuclide in the body or organs measured t days after intake (Bq), or the daily excretion (Bqd): m(t) is the expected value of the content of the nuclide in the body or organs t days after intake of a unit activity (Bq), or the daily excretion (Bqd\). The activity here can also be expressed as a fraction of the intake (except for xenon water). The m(t) values of some commonly used radionuclides under inhalation conditions are shown in Appendix A (Normative Appendix). The m(t) values when t>10 days can be found based on the relevant figures in Appendix A. If multiple measurement results are obtained, the intake can be estimated by the minimum multiplication method. 7.3 Dose calculation and evaluation
7.3.1 The intake is multiplied by the dose coefficient to obtain the accumulated effective dose. The dose can be evaluated by comparing the dose calculation results with the annual dose limit. The time-integrated air activity concentration obtained by PAS is multiplied by the volume of air inhaled by the worker during the intake period to obtain the intake of radionuclides. 7.3.2 When informing individual workers of the possible health effects of the monitoring results, the actual age at the time of intake should be considered. When conducting protection assessments, the intake can be directly compared with the annual intake limit (ALI). 7.4 Mixtures of multiple radionuclides
7.4.1 In the case of ingestion of a mixture of multiple radionuclides, generally only a few nuclides contribute significantly to the treated effective dose. In principle, it is necessary to first confirm which nuclides are of important radiobiological significance, and then formulate a monitoring plan for these nuclides.
7.4.2 When the composition of a mixture of multiple radionuclides is known and remains unchanged, nuclides with known metabolic laws and easy to measure but not necessarily important radiobiological significance can be used as "tracers" to infer the intake of other nuclides. 7.5 Investigation level and record level
7.5.1 The establishment of the investigation level is related to the purpose of the monitoring plan and the type of investigation to be conducted. For routine monitoring, different proportions of the annual dose limit or annual intake limit can be taken as the investigation level based on the understanding of workplace conditions and the specific circumstances. The same principle applies to the record level. 11
-ALI establishes the investigation level, and the monitoring period is T days. The derived investigation level DIL for routine monitoring is: 7.5.2 If 1
DIL = 0.1ALIX
In the formula:
365-number of days in a year;
T-monitoring period, days;
m(T/2)-same meaning as formula (1).
×mT/2)
When the measurement result exceeds the DIL, further investigation should be carried out. ....(3)
=ALI to establish the record level, the monitoring period is T days, then the derived record level DRL for routine monitoring7.5.3 If 1
ALI×365
xm（T/2）
where: 365, T and m（T/2）have the same meaning as in formula (3). Measurement results exceeding DRL should be recorded in the personal exposure record. (4)
7.5.4 Workers may be exposed to internal and external radiation, or mixed radionuclides at the same time, and this should be taken into account when establishing the investigation level and record level in advance.
8 Uncertainty and quality assurance of internal exposure monitoring8.1 Uncertainty
8.1.1 The uncertainty of dose estimation is the combination of the uncertainty components of the three stages of personal monitoring measurement, estimation of intake from measurement results, and estimation of dose from intake.
8.1.2 The uncertainty in measurement is generally easier to estimate. When the activity level is close to the detection limit, the uncertainty caused by counting statistical fluctuations is the main one. For radionuclides that are easy to measure and have a large enough activity, the uncertainty caused by counting statistical fluctuations is relatively small compared with other sources of uncertainty. In addition, the uncertainty introduced by other systems in the measurement process (such as calibration, correction of body size in direct measurement, etc.) and the errors caused by sample and body surface contamination must also be considered. 8.1.3 The biokinetic model of the radionuclide series recommended by ICRP is used to estimate the intake. The reliability of this estimate is related to the accuracy of the biokinetic model of radionuclides and its limitations in application under special circumstances. In the case of using ovulation-promoting drugs, this standard biokinetic model cannot be used to estimate the intake. 8.1.4 There is uncertainty in the process of estimating dose from a given intake. For routine monitoring, when the intake is within the annual intake limit, the default parameters of the standard biokinetic model can be used to estimate the intake accurately enough; for exposures that reach or exceed the annual intake limit, more detailed information on the physicochemical properties of the intake substance and the biokinetic parameters of the individual intake are required to improve the accuracy of the model estimate. 8.1.5 It is difficult to estimate the uncertainty of the intake estimate, so it is advisable to first estimate the intake based on a standard model and then2 Intakes generated in previous monitoring cycles may affect the measurement results of the current monitoring cycle. If more than about 10% of the current measurement value comes from intakes in previous monitoring cycles and its intake and dose have been estimated, the measurement results of the current monitoring cycle should be corrected. For a series of measurements in a routine monitoring plan, the following steps can be followed: a) determine the intake value of the first monitoring cycle; b) estimate the contribution of this intake to the measurement results of subsequent monitoring cycles; c) subtract this contribution from the data of subsequent monitoring cycles; d) repeat a) to c) for the next monitoring cycle. 7.1.3 In a routine monitoring plan, if the monitoring results exceed the predetermined investigation level, further investigation should be carried out. The nature of the investigation will depend on the specific circumstances and the extent to which the monitoring results exceed the investigation level. In the investigation, the following points should be considered: a) Repeat measurements to confirm or improve the initial assessment; b) Use additional monitoring techniques:
c) Evaluate working conditions and exposure situations;
d) If default parameter values were used in the initial assessment, if necessary, investigate the actual particle size and chemical form of the contaminants and select more appropriate values: e) In the case of large intakes, remove the contaminated person from radioactive work and monitor the retention and excretion characteristics of the contaminants in the body of the intake person to improve the dose assessment. 7.2 Special monitoring and task-related monitoring
In this case, the time of intake is known. If only one measurement is made, the intake I can be calculated by the following formula: I = M/m(t)
Where:
I is the intake of radionuclides, in becquerels (Bq): ·(2)
M is the content of the nuclide in the body or organs measured t days after intake (Bq), or the daily excretion (Bqd): m(t) is the expected value of the content of the nuclide in the body or organs t days after intake of a unit activity (Bq), or the daily excretion (Bqd\). The activity here can also be expressed as a fraction of the intake (except for xenon water). The m(t) values of some commonly used radionuclides under inhalation conditions are shown in Appendix A (Normative Appendix). The m(t) values when t>10 days can be found based on the relevant figures in Appendix A. If multiple measurement results are obtained, the intake can be estimated by the minimum multiplication method. 7.3 Dose calculation and evaluation
7.3.1 The intake is multiplied by the dose coefficient to obtain the accumulated effective dose. The dose can be evaluated by comparing the dose calculation results with the annual dose limit. The time-integrated air activity concentration obtained by PAS is multiplied by the volume of air inhaled by the worker during the intake period to obtain the intake of radionuclides. 7.3.2 When informing individual workers of the possible health effects of the monitoring results, the actual age at the time of intake should be considered. When conducting protection assessments, the intake can be directly compared with the annual intake limit (ALI). 7.4 Mixtures of multiple radionuclides
7.4.1 In the case of ingestion of a mixture of multiple radionuclides, generally only a few nuclides contribute significantly to the treated effective dose. In principle, it is necessary to first confirm which nuclides are of important radiobiological significance, and then formulate a monitoring plan for these nuclides.
7.4.2 When the composition of a mixture of multiple radionuclides is known and remains unchanged, nuclides with known metabolic laws and easy to measure but not necessarily important radiobiological significance can be used as "tracers" to infer the intake of other nuclides. 7.5 Investigation level and record level
7.5.1 The establishment of the investigation level is related to the purpose of the monitoring plan and the type of investigation to be conducted. For routine monitoring, different proportions of the annual dose limit or annual intake limit can be taken as the investigation level based on the understanding of workplace conditions and the specific circumstances. The same principle applies to the record level. 11
-ALI establishes the investigation level, and the monitoring period is T days. The derived investigation level DIL for routine monitoring is: 7.5.2 If 1
DIL = 0.1ALIX
In the formula:
365-number of days in a year;
T-monitoring period, days;
m(T/2)-same meaning as formula (1).
×mT/2)
When the measurement result exceeds the DIL, further investigation should be carried out. ....(3)
=ALI to establish the record level, the monitoring period is T days, then the derived record level DRL for routine monitoring7.5.3 If 1
ALI×365
xm（T/2）
where: 365, T and m（T/2）have the same meaning as in formula (3). Measurement results exceeding DRL should be recorded in the personal exposure record. (4)
7.5.4 Workers may be exposed to internal and external radiation, or mixed radionuclides at the same time, and this should be taken into account when establishing the investigation level and record level in advance.
8 Uncertainty and quality assurance of internal exposure monitoring8.1 Uncertainty
8.1.1 The uncertainty of dose estimation is the combination of the uncertainty components of the three stages of personal monitoring measurement, estimation of intake from measurement results, and estimation of dose from intake.
8.1.2 The uncertainty in measurement is generally easier to estimate. When the activity level is close to the detection limit, the uncertainty caused by counting statistical fluctuations is the main one. For radionuclides that are easy to measure and have a large enough activity, the uncertainty caused by counting statistical fluctuations is relatively small compared with other sources of uncertainty. In addition, the uncertainty introduced by other systems in the measurement process (such as calibration, correction of body size in direct measurement, etc.) and the errors caused by sample and body surface contamination must also be considered. 8.1.3 The biokinetic model of the radionuclide series recommended by ICRP is used to estimate the intake. The reliability of this estimate is related to the accuracy of the biokinetic model of radionuclides and its limitations in application under special circumstances. In the case of using ovulation-promoting drugs, this standard biokinetic model cannot be used to estimate the intake. 8.1.4 There is uncertainty in the process of estimating dose from a given intake. For routine monitoring, when the intake is within the annual intake limit, the default parameters of the standard biokinetic model can be used to estimate the intake accurately enough; for exposures that reach or exceed the annual intake limit, more detailed information on the physicochemical properties of the intake substance and the biokinetic parameters of the individual intake are required to improve the accuracy of the model estimate. 8.1.5 It is difficult to estimate the uncertainty of the intake estimate, so it is advisable to first estimate the intake based on a standard model and then2 Intakes generated in previous monitoring cycles may affect the measurement results of the current monitoring cycle. If more than about 10% of the current measurement value comes from intakes in previous monitoring cycles and its intake and dose have been estimated, the measurement results of the current monitoring cycle should be corrected. For a series of measurements in a routine monitoring plan, the following steps can be followed: a) determine the intake value of the first monitoring cycle; b) estimate the contribution of this intake to the measurement results of subsequent monitoring cycles; c) subtract this contribution from the data of subsequent monitoring cycles; d) repeat a) to c) for the next monitoring cycle. 7.1.3 In a routine monitoring plan, if the monitoring results exceed the predetermined investigation level, further investigation should be carried out. The nature of the investigation will depend on the specific circumstances and the extent to which the monitoring results exceed the investigation level. In the investigation, the following points should be considered: a) Repeat measurements to confirm or improve the initial assessment; b) Use additional monitoring techniques:
c) Evaluate working conditions and exposure situations;
d) If default parameter values were used in the initial assessment, if necessary, investigate the actual particle size and chemical form of the contaminants and select more appropriate values: e) In the case of large intakes, remove the contaminated person from radioactive work and monitor the retention and excretion characteristics of the contaminants in the body of the intake person to improve the dose assessment. 7.2 Special monitoring and task-related monitoring
In this case, the time of intake is known. If only one measurement is made, the intake I can be calculated by the following formula: I = M/m(t)
Where:
I is the intake of radionuclides, in becquerels (Bq): ·(2)
M is the content of the nuclide in the body or organs measured t days after intake (Bq), or the daily excretion (Bqd): m(t) is the expected value of the content of the nuclide in the body or organs t days after intake of a unit activity (Bq), or the daily excretion (Bqd\). The activity here can also be expressed as a fraction of the intake (except for xenon water). The m(t) values of some commonly used radionuclides under inhalation conditions are shown in Appendix A (Normative Appendix). The m(t) values when t>10 days can be found based on the relevant figures in Appendix A. If multiple measurement results are obtained, the intake can be estimated by the minimum multiplication method. 7.3 Dose calculation and evaluation
7.3.1 The intake is multiplied by the dose coefficient to obtain the accumulated effective dose. The dose can be evaluated by comparing the dose calculation results with the annual dose limit. The time-integrated air activity concentration obtained by PAS is multiplied by the volume of air inhaled by the worker during the intake period to obtain the intake of radionuclides. 7.3.2 When informing individual workers of the possible health effects of the monitoring results, the actual age at the time of intake should be considered. When conducting protection assessments, the intake can be directly compared with the annual intake limit (ALI). 7.4 Mixtures of multiple radionuclides
7.4.1 In the case of ingestion of a mixture of multiple radionuclides, generally only a few nuclides contribute significantly to the treated effective dose. In principle, it is necessary to first confirm which nuclides are of important radiobiological significance, and then formulate a monitoring plan for these nuclides.
7.4.2 When the composition of a mixture of multiple radionuclides is known and remains unchanged, nuclides with known metabolic laws and easy to measure but not necessarily important radiobiological significance can be used as "tracers" to infer the intake of other nuclides. 7.5 Investigation level and record level
7.5.1 The establishment of the investigation level is related to the purpose of the monitoring plan and the type of investigation to be conducted. For routine monitoring, different proportions of the annual dose limit or annual intake limit can be taken as the investigation level based on the understanding of workplace conditions and the specific circumstances. The same principle applies to the record level. 11
-ALI establishes the investigation level, and the monitoring period is T days. The derived investigation level DIL for routine monitoring is: 7.5.2 If 1
DIL = 0.1ALIX
In the formula:
365-number of days in a year;
T-monitoring period, days;
m(T/2)-same meaning as formula (1).
×mT/2)
When the measurement result exceeds the DIL, further investigation should be carried out. ....(3)
=ALI to establish the record level, the monitoring period is T days, then the derived record level DRL for routine monitoring7.5.3 If 1
ALI×365
xm（T/2）
where: 365, T and m（T/2）have the same meaning as in formula (3). Measurement results exceeding DRL should be recorded in the personal exposure record. (4)Www.bzxZ.net
7.5.4 Workers may be exposed to internal and external radiation, or mixed radionuclides at the same time, and this should be taken into account when establishing the investigation level and record level in advance.
8 Uncertainty and quality assurance of internal exposure monitoring8.1 Uncertainty
8.1.1 The uncertainty of dose estimation is the combination of the uncertainty components of the three stages of personal monitoring measurement, estimation of intake from measurement results, and estimation of dose from intake.
8.1.2 The uncertainty in measurement is generally easier to estimate. When the activity level is close to the detection limit, the uncertainty caused by counting statistical fluctuations is the main one. For radionuclides that are easy to measure and have a large enough activity, the uncertainty caused by counting statistical fluctuations is relatively small compared with other sources of uncertainty. In addition, the uncertainty introduced by other systems in the measurement process (such as calibration, correction of body size in direct measurement, etc.) and the errors caused by sample and body surface contamination must also be considered. 8.1.3 The biokinetic model of the radionuclide series recommended by ICRP is used to estimate the intake. The reliability of this estimate is related to the accuracy of the biokinetic model of radionuclides and its limitations in application under special circumstances. In the case of using ovulation-promoting drugs, this standard biokinetic model cannot be used to estimate the intake. 8.1.4 There is uncertainty in the process of estimating dose from a given intake. For routine monitoring, when the intake is within the annual intake limit, the default parameters of the standard biokinetic model can be used to estimate the intake accurately enough; for exposures that reach or exceed the annual intake limit, more detailed information on the physicochemical properties of the intake substance and the biokinetic parameters of the individual intake are required to improve the accuracy of the model estimate. 8.1.5 It is difficult to estimate the uncertainty of the intake estimate, so it is advisable to first estimate the intake based on a standard model and then1 The intake is multiplied by the dose coefficient to obtain the committed effective dose. The dose can be evaluated by comparing the dose calculation results with the annual dose limit. The time-integrated air activity concentration obtained by PAS is multiplied by the volume of air inhaled by the worker during the intake period to obtain the intake of radionuclides. 7.3.2 When informing individual workers of the possible health effects of the monitoring results, the actual age at the time of intake should be considered. When conducting protection assessment, the intake can be directly compared with the annual intake limit (ALI). 7.4 Mixtures of multiple radionuclides
7.4.1 In the case of the intake of a mixture of multiple radionuclides, generally only a few nuclides contribute significantly to the committed effective dose. In principle, it is necessary to first confirm which nuclides are of important radiobiological significance, and then formulate a monitoring plan for these nuclides.
7.4.2 When the composition of a mixture of multiple radionuclides is known and remains unchanged, nuclides with known metabolic laws and easy to measure but not necessarily important radiobiological significance can be used as "tracers" to infer the intake of other nuclides. 7.5 Investigation level and record level
7.5.1 The establishment of the investigation level is related to the purpose of the monitoring plan and the type of investigation to be conducted. For routine monitoring, different proportions of the annual dose limit or annual intake limit can be taken as the investigation level based on the understanding of workplace conditions and the specific circumstances. The same principle applies to the record level. 11
-ALI establishes the investigation level, and the monitoring period is T days. The derived investigation level DIL for routine monitoring is: 7.5.2 If 1
DIL = 0.1ALIX
In the formula:
365-number of days in a year;
T-monitoring period, days;
m(T/2)-same meaning as formula (1).
×mT/2)
When the measurement result exceeds the DIL, further investigation should be carried out. ....(3)
=ALI to establish the record level, the monitoring period is T days, then the derived record level DRL for routine monitoring7.5.3 If 1
ALI×365
xm（T/2）
where: 365, T and m（T/2）have the same meaning as in formula (3). Measurement results exceeding DRL should be recorded in the personal exposure record. (4)
7.5.4 Workers may be exposed to internal and external radiation, or mixed radionuclides at the same time, and this should be taken into account when establishing the investigation level and record level in advance.
8 Uncertainty and quality assurance of internal exposure monitoring8.1 Uncertainty
8.1.1 The uncertainty of dose estimation is the combination of the uncertainty components of the three stages of personal monitoring measurement, estimation of intake from measurement results, and estimation of dose from intake.
8.1.2 The uncertainty in measurement is generally easier to estimate. When the activity level is close to the detection limit, the uncertainty caused by counting statistical fluctuations is the main one. For radionuclides that are easy to measure and have a large enough activity, the uncertainty caused by counting statistical fluctuations is relatively small compared with other sources of uncertainty. In addition, the uncertainty introduced by other systems in the measurement process (such as calibration, correction of body size in direct measurement, etc.) and the errors caused by sample and body surface contamination must also be considered. 8.1.3 The biokinetic model of the radionuclide series recommended by ICRP is used to estimate the intake. The reliability of this estimate is related to the accuracy of the biokinetic model of radionuclides and its limitations in application under special circumstances. In the case of using ovulation-promoting drugs, this standard biokinetic model cannot be used to estimate the intake. 8.1.4 There is uncertainty in the process of estimating dose from a given intake. For routine monitoring, when the intake is within the annual intake limit, the default parameters of the standard biokinetic model can be used to estimate the intake accurately enough; for exposures that reach or exceed the annual intake limit, more detailed information on the physicochemical properties of the intake substance and the biokinetic parameters of the individual intake are required to improve the accuracy of the model estimate. 8.1.5 It is difficult to estimate the uncertainty of the intake estimate, so it is advisable to first estimate the intake based on a standard model and then1 The intake is multiplied by the dose coefficient to obtain the committed effective dose. The dose can be evaluated by comparing the dose calculation results with the annual dose limit. The time-integrated air activity concentration obtained by PAS is multiplied by the volume of air inhaled by the worker during the intake period to obtain the intake of radionuclides. 7.3.2 When informing individual workers of the possible health effects of the monitoring results, the actual age at the time of intake should be considered. When conducting protection assessment, the intake can be directly compared with the annual intake limit (ALI). 7.4 Mixtures of multiple radionuclides
7.4.1 In the case of the intake of a mixture of multiple radionuclides, generally only a few nuclides contribute significantly to the committed effective dose. In principle, it is necessary to first confirm which nuclides are of important radiobiological significance, and then formulate a monitoring plan for these nuclides.
7.4.2 When the composition of a mixture of multiple radionuclides is known and remains unchanged, nuclides with known metabolic laws and easy to measure but not necessarily important radiobiological significance can be used as "tracers" to infer the intake of other nuclides. 7.5 Investigation level and record level
7.5.1 The establishment of the investigation level is related to the purpose of the monitoring plan and the type of investigation to be conducted. For routine monitoring, different proportions of the annual dose limit or annual intake limit can be taken as the investigation level based on the understanding of workplace conditions and the specific circumstances. The same principle applies to the record level. 11
-ALI establishes the investigation level, and the monitoring period is T days. The derived investigation level DIL for routine monitoring is: 7.5.2 If 1
DIL = 0.1ALIX
In the formula:
365-number of days in a year;
T-monitoring period, days;
m(T/2)-same meaning as formula (1).
×mT/2)
When the measurement result exceeds the DIL, further investigation should be carried out. ....(3)
=ALI to establish the record level, the monitoring period is T days, then the derived record level DRL for routine monitoring7.5.3 If 1
ALI×365
xm（T/2）
where: 365, T and m（T/2）have the same meaning as in formula (3). Measurement results exceeding DRL should be recorded in the personal exposure record. (4)
7.5.4 Workers may be exposed to internal and external radiation, or mixed radionuclides at the same time, and this should be taken into account when establishing the investigation level and record level in advance.
8 Uncertainty and quality assurance of internal exposure monitoring8.1 Uncertainty
8.1.1 The uncertainty of dose estimation is the combination of the uncertainty components of the three stages of personal monitoring measurement, estimation of intake from measurement results, and estimation of dose from intake.
8.1.2 The uncertainty in measurement is generally easier to estimate. When the activity level is close to the detection limit, the uncertainty caused by counting statistical fluctuations is the main one. For radionuclides that are easy to measure and have a large enough activity, the uncertainty caused by counting statistical fluctuations is relatively small compared with other sources of uncertainty. In addition, the uncertainty introduced by other systems in the measurement process (such as calibration, correction of body size in direct measurement, etc.) and the errors caused by sample and body surface contamination must also be considered. 8.1.3 The biokinetic model of the radionuclide series recommended by ICRP is used to estimate the intake. The reliability of this estimate is related to the accuracy of the biokinetic model of radionuclides and its limitations in application under special circumstances. In the case of using ovulation-promoting drugs, this standard biokinetic model cannot be used to estimate the intake. 8.1.4 There is uncertainty in the process of estimating dose from a given intake. For routine monitoring, when the intake is within the annual intake limit, the default parameters of the standard biokinetic model can be used to estimate the intake accurately enough; for exposures that reach or exceed the annual intake limit, more detailed information on the physicochemical properties of the intake substance and the biokinetic parameters of the individual intake are required to improve the accuracy of the model estimate. 8.1.5 It is difficult to estimate the uncertainty of the intake estimate, so it is advisable to first estimate the intake based on a standard model and then
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