This standard specifies the requirements and methods for the characterisation of low-level radioactive waste packages with cement as the main curing (fixation) medium. This standard applies to the characterisation of low-level radioactive waste packages (both solidified and immobilised) for near-surface disposal. This standard does not apply to the characterisation of waste packages in high integrity containers.
2 Normative references
The following documents or clauses from them are cited in this standard. Where references are dated, only the dated version applies to this standard. Where a document is referenced without a date, the latest version (including all amendment sheets) applies to this standard.
3 Terminology and definitions
The following terms and definitions apply to this standard.
4 Waste package characterisation requirements
4.1 General requirements
4.1.1 Waste package characterisation is appropriate according to the waste source, waste stream characteristics, waste classification and collection, treatment, preparation, storage, transport and other processes of the characteristics of data investigation, analysis, the results obtained is one of the basis for the evaluation of the performance of the waste package.
4.1.2 Waste package characterisation can be carried out through documentary inspection, visual inspection and direct inspection means. That is, by checking the documentation provided by the waste-generating unit to determine whether the waste package meets the requirements of GB9132 and GB 12711; by checking whether the surface of the waste package is broken, cracked, deformed, the lid of the drum (box) is locked, and whether the logo, code is clear, etc., to determine whether the requirements of the waste reception standards are met; you can also take a certain amount of waste packages in a certain proportion for direct measurement.
4.1.3 In the waste treatment and preparation phase, a sampling plan needs to be developed based on the waste source item, treatment and/or preparation process, and sampling analysis needs to be carried out to determine the waste components and radioactive properties. In the waste formation process, if it has been demonstrated through bypass sampling or laboratory simulations that the waste properties can meet the performance requirements specified in 4.2.1, then the waste source item, treatment and preparation process and process parameters remain unchanged, the resulting waste object is considered to meet the requirements of the standard, and the waste package performance test can no longer be carried out; if the waste source item, treatment and preparation process and process parameters change significantly, then the performance test must be carried out. If the waste source item, treatment process and process parameters change significantly, a performance test must be conducted to verify that the resulting waste object/waste package meets the requirements of the standard.
4.1.4 The radionuclide composition and activity concentration of the waste package are central to the characterisation of the waste package. Measurements of the waste source term are required before the waste object is formed; after the waste object is formed, verification of the radiological properties is appropriate. The measurement method can be determined according to the waste source pathway and whether the radionuclides are homogeneously distributed in the waste. To simplify measurements, characterisation methods for wastes of similar origin can be established by the test-model-validation route. The parameters, models as well as methods used should be credible and validated or proven.
4.2 Waste package performance and testing requirements
4.2.2 Waste containers
Waste containers should be steel containers meeting the requirements of EJ 1042, EJ1076 or concrete containers meeting the requirements of EJ914.
4.2.3 Waste packages
4.2.3.1 The type and activity concentration of radionuclides in the waste package shall comply with the requirements of the limits for low-level radioactive waste in the Classification of Radioactive Waste.
4.2.3.2 The total activity of dense in a single dense waste package shall not exceed 1.3 x 103 Bq and the dense release rate shall not be greater than 1 part in 100,000 of the total activity per month.
4.2.3.3 Limit values for surface contamination of waste packages:
β, y emitters, low toxicity alpha emitters ≤ 4 Bq/cm2 other alpha emitters ≤ 0.4 Bq/cm2
4.2.3.4 The content of free liquid in the waste package should not exceed 1% of the waste volume.
4.2.3.5 The fill rate of the waste package should be no less than 90% of the full volume of the waste container.
4.3 Waste package inspection items specified
Waste package formation process, test items, test methods and test frequency should be in accordance with the provisions of Table 2, Table 3.
See Appendix A for the principles of selection of radionuclides to be measured.
5.1.2 Direct measurement method
Sampling methods can be used to determine the radionuclide composition and activity of the waste/waste package using energy spectrometry and/or radiochemical analysis by reference to Appendix B.
5.1.3 Non-destructive measurements
5.1.3.1 Non-destructive measurements such as y-rays, neutrons, etc., using validated and qualified procedures, models and parameters such as the waste source item information and waste object generation process, waste container shielding parameters, distribution of nuclides in the waste object, etc., may be used to derive the composition and activity of radionuclides in the waste object.
5.1.3.2 Non-destructive measurement methods should be validated during the build-up phase and subjected to uncertainty analysis; they should be validated for reasonableness when extending their use to different scenarios; and important parameters of the model should also be validated periodically during the use phase.
5.1.4 Key nuclide extrapolation method
If the quantitative relationships between the different radionuclides in the waste are known and there is at least one key nuclide that can be easily measured, the amount of the other nuclides and the activity of the waste material can be calculated from the measurements of that key nuclide using conversion factors or equilibrium relationships between the nuclides; the equilibrium relationships and conversion factors between the nuclides should be validated periodically during the use phase.
5.1.5 Calculation method
Calculation or estimation of radionuclides and their specific activity in the waste object by means of the composition of radionuclides in the waste and/or the generation process and its key parameters, including e.g. neutron injection rate, irradiation time, average fuel consumption, cooling decay time, material-radioactivity equilibrium calculations, calculated parameters for migration of nuclides in the process and the volume of the waste package in the solidified body, etc., using validated procedures.
5.2 Chemical performance test methods
5.3 Physical properties test methods
5.3.1 Homogeneity
The homogeneity of the waste object is preferably determined by chromatographic radiography, but may also be determined by dissection, etc.
5.3.2 Free liquids
5.3.2.1 Free liquids in cement curing body inspection, should be in line with GB 14569.1 provisions.
5.3.2.2 Free liquids in the waste fixation body can be determined by radiography, laminar X-ray radiography or ultrasound.
5.3.3 Impermeability of the fixation medium
Cement mortar and fine stone concrete shall be tested for impermeability in accordance with Appendix C of EJ914.
5.3.4 Flow rate
5.3.4.1 Determination of cement mortar flow, should be consistent with the provisions of GB/T2419.
5.3.4.2 Determination of the collapse expansion of fine concrete, should be consistent with the provisions of GB/T 50080.
5.4 Mechanical properties test methods
5.4.1 cement curing body compressive strength, impact resistance test, should be consistent with the provisions of GB 14569.1.
5.4.2 Cement mortar compressive strength test, should be consistent with the provisions of GB/T 17671.
5.4.3 fine stone concrete compressive strength test, should be consistent with the provisions of GB/T 50081.
5.5 irradiation stability test method
Cement curing body resistance to y irradiation test, should be consistent with the provisions of GB 14569.1 in 6.5.
6.1.1 Waste packages should be based on the waste source items and/or generated batches, according to the proportion of not more than 2% of each batch, at least two pieces of sampling. The random inspection is based on a non-destructive test and, where necessary, samples may be taken for analysis by reference to the provisions of Appendix B, including some or all of the following:
a) the appearance of the waste package;
b) the quality of the waste package;
c) the surface contamination level of the waste package;
d) the surface radiation dose level of the waste package;
e) the type and activity of radionuclides in the waste package;
f) the void and fill rate of the waste package;
g) the type and nature of the contents of the waste package.
6.1.2 For waste packages used directly as transport packages without the use of transport containers, it should also be confirmed that the requirements of GB 11806 are met. 6.1.3 Physical inspection of the waste package may be carried out using a device in accordance with GB/T 19211.
6.2 Radioactivity measurements
6.2.1 Dose rate measurements
7 Quality assurance
7.1 Units related to the characterization of waste objects and waste packages should prepare quality assurance documents for characterization, including organizational structure, staff and equipment configuration, management procedures, technical requirements, etc. Technical requirements for characterisation as well as assurance measures should be specified.
7.2 The waste generating unit should carry out quality control and assessment of each process, process equipment and instrumentation to ensure that the production process is under control when handling the whole waste.
Appendix A (Informative Appendix) Nucleotide Selection
1 Scope of application
This standard specifies the requirements and methods for the characterisation of low-level radioactive waste packages with cement as the main curing (fixation) medium. This standard applies to the characterisation of low-level radioactive waste packages (both solidified and immobilised) for near-surface disposal. This standard does not apply to the characterisation of waste packages in high integrity containers.
2 Normative references
The following documents or clauses from them are cited in this standard. Where references are dated, only the dated version applies to this standard. Where a document is referenced without a date, the latest version (including all amendment sheets) applies to this standard.
3 Terminology and definitions
The following terms and definitions apply to this standard.
4 Waste package characterisation requirements
4.1 General requirements
4.1.1 Waste package characterisation is appropriate according to the waste source, waste stream characteristics, waste classification and collection, treatment, preparation, storage, transport and other processes of the characteristics of data investigation, analysis, the results obtained is one of the basis for the evaluation of the performance of the waste package.
4.1.2 Waste package characterisation can be carried out through documentary inspection, visual inspection and direct inspection means. That is, by checking the documentation provided by the waste-generating unit to determine whether the waste package meets the requirements of GB9132 and GB 12711; by checking whether the surface of the waste package is broken, cracked, deformed, the lid of the drum (box) is locked, and whether the logo, code is clear, etc., to determine whether the requirements of the waste reception standards are met; you can also take a certain amount of waste packages in a certain proportion for direct measurement.
4.1.3 In the waste treatment and preparation phase, a sampling plan needs to be developed based on the waste source item, treatment and/or preparation process, and sampling analysis needs to be carried out to determine the waste components and radioactive properties. In the waste formation process, if it has been demonstrated through bypass sampling or laboratory simulations that the waste properties can meet the performance requirements specified in 4.2.1, then the waste source item, treatment and preparation process and process parameters remain unchanged, the resulting waste object is considered to meet the requirements of the standard, and the waste package performance test can no longer be carried out; if the waste source item, treatment and preparation process and process parameters change significantly, then the performance test must be carried out. If the waste source item, treatment process and process parameters change significantly, a performance test must be conducted to verify that the resulting waste object/waste package meets the requirements of the standard.
4.1.4 The radionuclide composition and activity concentration of the waste package are central to the characterisation of the waste package. Measurements of the waste source term are required before the waste object is formed; after the waste object is formed, verification of the radiological properties is appropriate. The measurement method can be determined according to the waste source pathway and whether the radionuclides are homogeneously distributed in the waste. To simplify measurements, characterisation methods for wastes of similar origin can be established by the test-model-validation route. The parameters, models as well as methods used should be credible and validated or proven.
4.2 Waste package performance and testing requirements
4.2.2 Waste containers
Waste containers should be steel containers meeting the requirements of EJ 1042, EJ1076 or concrete containers meeting the requirements of EJ914.
4.2.3 Waste packages
4.2.3.1 The type and activity concentration of radionuclides in the waste package shall comply with the requirements of the limits for low-level radioactive waste in the Classification of Radioactive Waste.
4.2.3.2 The total activity of dense in a single dense waste package shall not exceed 1.3 x 103 Bq and the dense release rate shall not be greater than 1 part in 100,000 of the total activity per month.
4.2.3.3 Limit values for surface contamination of waste packages:
β, y emitters, low toxicity alpha emitters ≤ 4 Bq/cm2 other alpha emitters ≤ 0.4 Bq/cm2
4.2.3.4 The content of free liquid in the waste package should not exceed 1% of the waste volume.
4.2.3.5 The fill rate of the waste package should be no less than 90% of the full volume of the waste container.
4.3 Waste package inspection items specified
Waste package formation process, test items, test methods and test frequency should be in accordance with the provisions of Table 2, Table 3.
5 Waste object characteristics identification methods
5.1 Radioactivity measurement
5.1.1 Principle of nuclide selection
See Appendix A for the principles of selection of radionuclides to be measured.
5.1.2 Direct measurement method
Sampling methods can be used to determine the radionuclide composition and activity of the waste/waste package using energy spectrometry and/or radiochemical analysis by reference to Appendix B.
5.1.3 Non-destructive measurements
5.1.3.1 Non-destructive measurements such as y-rays, neutrons, etc., using validated and qualified procedures, models and parameters such as the waste source item information and waste object generation process, waste container shielding parameters, distribution of nuclides in the waste object, etc., may be used to derive the composition and activity of radionuclides in the waste object.
5.1.3.2 Non-destructive measurement methods should be validated during the build-up phase and subjected to uncertainty analysis; they should be validated for reasonableness when extending their use to different scenarios; and important parameters of the model should also be validated periodically during the use phase.
5.1.4 Key nuclide extrapolation method
If the quantitative relationships between the different radionuclides in the waste are known and there is at least one key nuclide that can be easily measured, the amount of the other nuclides and the activity of the waste material can be calculated from the measurements of that key nuclide using conversion factors or equilibrium relationships between the nuclides; the equilibrium relationships and conversion factors between the nuclides should be validated periodically during the use phase.
5.1.5 Calculation method
Calculation or estimation of radionuclides and their specific activity in the waste object by means of the composition of radionuclides in the waste and/or the generation process and its key parameters, including e.g. neutron injection rate, irradiation time, average fuel consumption, cooling decay time, material-radioactivity equilibrium calculations, calculated parameters for migration of nuclides in the process and the volume of the waste package in the solidified body, etc., using validated procedures.
5.2 Chemical performance test methods
5.3 Physical properties test methods
5.3.1 Homogeneity
The homogeneity of the waste object is preferably determined by chromatographic radiography, but may also be determined by dissection, etc.
5.3.2 Free liquids
5.3.2.1 Free liquids in cement curing body inspection, should be in line with GB 14569.1 provisions.
5.3.2.2 Free liquids in the waste fixation body can be determined by radiography, laminar X-ray radiography or ultrasound.
5.3.3 Impermeability of the fixation medium
Cement mortar and fine stone concrete shall be tested for impermeability in accordance with Appendix C of EJ914.
5.3.4 Flow rate
5.3.4.1 Determination of cement mortar flow, should be consistent with the provisions of GB/T2419.
5.3.4.2 Determination of the collapse expansion of fine concrete, should be consistent with the provisions of GB/T 50080.
5.4 Mechanical properties test methods
5.4.1 cement curing body compressive strength, impact resistance test, should be consistent with the provisions of GB 14569.1.
5.4.2 Cement mortar compressive strength test, should be consistent with the provisions of GB/T 17671.
5.4.3 fine stone concrete compressive strength test, should be consistent with the provisions of GB/T 50081.
5.5 irradiation stability test method
Cement curing body resistance to y irradiation test, should be consistent with the provisions of GB 14569.1 in 6.5.
6 waste package characteristics identification methods
6.1 Waste package physical inspection
6.1.1 Waste packages should be based on the waste source items and/or generated batches, according to the proportion of not more than 2% of each batch, at least two pieces of sampling. The random inspection is based on a non-destructive test and, where necessary, samples may be taken for analysis by reference to the provisions of Appendix B, including some or all of the following:
a) the appearance of the waste package;
b) the quality of the waste package;
c) the surface contamination level of the waste package;
d) the surface radiation dose level of the waste package;
e) the type and activity of radionuclides in the waste package;
f) the void and fill rate of the waste package;
g) the type and nature of the contents of the waste package.
6.1.2 For waste packages used directly as transport packages without the use of transport containers, it should also be confirmed that the requirements of GB 11806 are met. 6.1.3 Physical inspection of the waste package may be carried out using a device in accordance with GB/T 19211.
6.2 Radioactivity measurements
6.2.1 Dose rate measurements
7 Quality assurance
7.1 Units related to the characterization of waste objects and waste packages should prepare quality assurance documents for characterization, including organizational structure, staff and equipment configuration, management procedures, technical requirements, etc. Technical requirements for characterisation as well as assurance measures should be specified.
7.2 The waste generating unit should carry out quality control and assessment of each process, process equipment and instrumentation to ensure that the production process is under control when handling the whole waste.
Appendix A (Informative Appendix) Nucleotide Selection
Appendix B (Informative Appendix) Sampling
