RADIATION PROTECTION

Study on induced radioactivity of D-T neutron tube radiography device

WANG Zhenyu, HUANG Weiqi, LAI Yongfang, SUN Jian

RADIATION PROTECTION. 2021, 41(S1): 1-6.

Abstract ( 110 )

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This article introduces the application and advantages of the D-T neutron tube camera device, and analyzes its induced radioactive hazards in combination with the characteristics of the D-T neutron tube and the characteristics of fast neutron photography. The induced radioactivity was calculated using the ENDF database. Relying on the IAEA’s definition and calculation method of exempt activity, the concept of risk index was put forward to visually express the degree of danger of nuclear induced radioactivity. Using methods such as calculation of risk index and horizontal comparison, the contribution of common nuclides, elements, and stainless steel to induced radioactivity risk were compared. The study found that the elements Mo, Fe, and Cu have greater risks, and the ferritic stainless steel 16Cr25 N has less risks. In the selection of stainless steel, the content of Fe, Ni, Mn and Cr should be paid attention to.

Evaluation of measurement uncertainty in external exposure personal dose monitoring

LIANG Na, WANG Yue, YANG Lili

RADIATION PROTECTION. 2021, 41(S1): 7-11.

Abstract ( 109 )

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External radiation personal dose monitoring data can be used as the occupational radiation dose certification received by radioactive workers and has the legal effect. Therefore, the accuracy and reliability of personal dose monitoring and measurement data for external irradiation is particularly important. Personal Dose Monitoring Centre in CNNP Nuclear Power Operation and Management Co.Ltd., obtained CMA (China Inspection Body and Laboratory Mandatory Approval) and the qualification of radiological health technical service institution in 2017. In addition to the center’s internal quality control measures, Centre also attends the annual inspection detection capacity assessment of the national radiological technology agency. This paper analysis the measurement data and uncertainty evaluation based on the 2019 capacity assessment experimental results (qualified) of CNNP. The relative extended uncertainty of the measurement results of each experimental group is about 10%, which proves that the external radiation monitoring and measurement system has a good performance, and ensures the accuracy and reliability of the monitoring data of the individual dose monitoring center.

Implementation and effect analysis of radiation environmental quality assessment based on environmental monitoring data

LI Yang, KANG Jing, WANG Yan, GU Zhijie

RADIATION PROTECTION. 2021, 41(S1): 12-14.

Abstract ( 100 )

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In order to ensure the implementation of the assessment of the radiation environmental quality status of nuclear bases and nuclear facilities in China, a radiation environmental quality method based on environmental monitoring data was established. The assessment method was introduced firstly, the implementation effect and problems encountered were analysed, and suggestions for improvement were put forward in the end.

Calculation of radioactivity with gaseous releases and liquid discharges from VVER-1000 units on normal operations

ZHANG Junnan, ZHOU Yaoquan, LI Lu, ZHENG Wei

RADIATION PROTECTION. 2021, 41(S1): 15-19.

Abstract ( 82 )

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VVER-1000 units was designed and manipulated by Russian in Tianwan nuclear power station, of which radioactivity with gaseous and liquid discharges on normal operations is one of the most important index that can testify the design of NPP satisfy the relative national environmental standard’s requirements, and one of the most important content of radiation protection optimization. This paper is based on the national pressurized water reactor NPP source term framework, by analyzing Tianwan NPP relative system process flow, established mathematical models of each equipment in different circuit system when the reactor is under normal operations, adopted design and operational parameters, and calculated realistic and conservative source term respectively. The results is compared with Russian’s, and presents the changes after the new source term framework being adopted.

The study on field measurement technology of beta surface contamination

LI Yuqin, WEN Fuping, LU Ying

RADIATION PROTECTION. 2021, 41(S1): 20-25.

Abstract ( 105 )

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In order to explore the influence of different influencing factors on the measurement efficiency of beta surface contamination, this paper mainly uses CoMo170 surface contamination monitor to measure the ⁶⁰Co plane source, the ²⁰⁴Tl plane source and the ⁹⁰Sr-⁹⁰Y plane source, and studies the influence of energy response, detection window response uniformity, measurement spacing, absorption effect, gamma ray interference and back scattering on the surface activity response, and the measurement uncertainty is evaluated. Through experimental research, the influence on the accuracy of the measurement results is obtained. Among them, the most influential factors are energy response and gamma ray interference. The final uncertainty evaluation result of the experimental measurement shows that its relative synthetic standard uncertainty is approximately equal to 46.71%.

Study on protection of workers from β-irradiation in the process of uranium purification and transformation

MA Wencai, LIU Yanzhang, ZHAO Ying

RADIATION PROTECTION. 2021, 41(S1): 26-28.

Abstract ( 78 )

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In order to strengthen the protection of workers from β irradiation in the process of uranium purification and transformation,combined with the practice of radiation monitoring at home and abroad,the characteristic of radiation source term in the process of uranium purification and transformation are studied. Due to the existence of main β emitter of Th-234,Pa-234,Bi-214 and Pb-214 in uranium system,the β radiation intensity in uranium compounds in considerable. Without protective measures,especially when carrying out open maintenance work,the exposure dose of β to lens、hands and skin of workers may exceed the prescribed limit. Therefore,in addition to the protection of external γradiation and internal inhalation radiation,it is necessary to protect workers from β irradiation.

Monitoring and evaluation of air activation in China spallation neutron source

WU Qingbiao, ZHUANG Sixuan

RADIATION PROTECTION. 2021, 41(S1): 29-35.

Abstract ( 116 )

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China Spallation Neutron Source (CSNS) is a large scientific experimental device of proton accelerator in operation in China. It is of great significance to ensure the normal operation and scientific research output of the equipment and the radiation safety of the staff and the public, and to ensure the air activation level at a monitored and controllable state. The previous understandings of accelerator air activation at home and abroad are almost coming from references and theoretical calculations. Due to lack of specific research and design, no air induced radionuclides and their concentrations have been measured in previous domestic accelerators’ air activation monitoring. Late of 2019, an online air activation monitoring system was developed in CSNS, which successfully detected the major radionuclides and their activity concentrations in air activation. This paper introduces the research and development of the system and the monitoring results, and puts forward some views on the monitoring and evaluation of accelerator air activation.

Proficiency testing in the field of radioactive measurement

WANG Ruijun, BAO Li, LI Pengxiang, LI Zhou, SONG Qinnan

RADIATION PROTECTION. 2021, 41(S1): 36-39.

Abstract ( 91 )

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There has been no proficiency testing provider in the field of radioactive measurement for a long time, and the personnel engaged in radioactive measurement do not have a deep understanding of inter laboratory comparison and proficiency testing. The key process of proficiency testing includes the preparation of proficiency testing sample, uniformity and stability inspection of proficiency testing items, packaging and distribution, evaluation of participants’ results, etc. The assessment of participants’ results can refer to the assessment method of IAEA, which is more comprehensive and reliable.

Development of protective clothing for external irradiation

WANG Wenxiu, CHEN Xiaojun, YIN Min, YANG Xiuyu

RADIATION PROTECTION. 2021, 41(S1): 40-44.

Abstract ( 87 )

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In this paper, the shielding material for shielding neutrons and gamma rays with certain energy are studied by using MCNP software. The shielding material is used to make protective vest for neutron and gamma radiation. The armor is thick and heavy, and it is difficult for human body to bear its weight. By studying the power-assisted exoskeleton device, the power-assisted exoskeleton device and the protective vest were combined to form protective clothing,and the power-assisted exoskeleton device was used to bear the weight of the protective vest, thus solving the problem that the human body needed to bear the weight. The performance of the protective clothing was tested with a radioactive source and in the field, and the technical parameters of the protective clothing for external irradiation, such as the shielding rate, thickness, uniformity and technical parameters have reached the expected requirements.

Some discussions about Asia-Pacific ALMERA group proficiency test and inter-comparison exercise on the determination of gross-alpha and gross-beta in filter

ZHANG Jing, LI Pengxiang, MA Xuyuan, LI Zhou, BAO Li, LIANG Runcheng

RADIATION PROTECTION. 2021, 41(S1): 45-49.

Abstract ( 79 )

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The 2015 Asia-Pacific Analytical Laboratories for the Measurement of Environmental Radioactivity (ALMERA) group proficiency test (PT) on the determination of Gross-alpha and Gross-beta and inter-comparison exercise on the gamma emitting radionuclides were conducted by IAEA, and participated in the PT of 14 ALMERA laboratories from 11 countries. The determination of Gross-alpha and Gross-beta is simpler than that of gamma emitting radionuclides. Gross-alpha and Gross-beta is measured separately, and there is no need to consider the mutual interference when measuring at the same time.Only 9 laboratories reported data with pass rate of 67% about the determination of Gross-alpha, and 10 laboratories reported data with pass rate of 50%.It is worthy of attention by those working in the survey.The article also introduced the detailed process of this experiment participating in this comparison. Through the direct measurement method of aerosol filter membranes, the relative deviations of total α and total β were 2.50% and -3.95%, respectively, and our laboratory successfully passed the comparison and summed up the experience from it.

Based on accident process to estimate fission number of uranyl fluoride solution criticality

JIA Linsheng, ZHANG Jiangang, YANG Yapeng, WANG Renze, FENG Zongyang, WANG Ning, LIANG Boning

RADIATION PROTECTION. 2021, 41(S1): 50-54.

Abstract ( 72 )

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Uranyl fluoride solution criticality accident is one of potential criticality accidents for nuclear fuel cycle facility. It is necessary to make corresponding accident emergency assessment, which can provide auxiliary support for emergency decision. Fission number of the criticality is an important part and one of the technical difficulties in nuclear critical accidents emergency assessment. It reflects the size and scale of nuclear criticality accident, and directly affects the decision of emergency protection action. There are many ways to estimate the fission number, and have themselves applicable conditions. With the time passed, the more abundant information obtained, and the selected evaluation method may change to follow it. Therefore, a method for estimating fission number of uranyl fluoride solution criticality based on accident process was proposed, which solved the practical application of critical accident emergency assessment and the difficult problem of which method users should choose.

Source term estimation of organic phase fire accident of 1A column in nuclear fuel reprocessing plants

LIANG Boning, ZHANG Jiangang, WANG Renze, YANG Yapeng, FENG Zongyang, JIA Linsheng, WANG Ning, LIU Yining

RADIATION PROTECTION. 2021, 41(S1): 55-58.

Abstract ( 76 )

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With the rapid growth of nuclear power industry in China, the quest for nuclear fuel reprocessing industry is getting stronger. In order to satisfy the quest for the nuclear emergency which is necessary for the safety development of nuclear fuel reprocessing industry and supply the evidence for nuclear emergency response and decision in nuclear fuel reprocessing plants. This paper aims at organic phase fire accident of 1A column which is a design basis accident in nuclear fuel reprocessing plants. Combined with the actual process flow and monitoring measures, this paper selects the available parameters as the input. Combined with the process characteristics of nuclear fuel reprocessing and minor corrections, this paper establishes the model for the organic phase fire accident of 1A column based on the empirical formula of organic phase burning rate. In addition, FORTRAN has been chosen to code the software. And the numerical results show that the estimation model can satisfy the quest of nuclear emergency during organic phase fire accident of 1A column in nuclear fuel reprocessing plants.

Research on fault diagnosis methods for nuclear power plant based on Naive Bayes

QI Ben, LIANG Jingang, ZHANG Liguo, TONG Jiejuan, YAN Shu

RADIATION PROTECTION. 2021, 41(S1): 59-63.

Abstract ( 83 )

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This paper introduces Bayesian techniques from the machine learning field into the application of power plant accident diagnosis in nuclear emergency. A new approach for plant accident diagnosis based on Naive Bayes Classifier is proposed. The PWR nuclear power plant simulator is used to obtain accident case data, and the naive Bayes classification model is trained to realize the diagnosis of multiple types of accidents (LOCA, SGTR, MSLB) in nuclear power plants. The test results show that the accident diagnosis methods based on naive Bayesian classifier have significant advantages in diagnostic accuracy, diagnostic efficiency, expandability of accident types and program autonomy diagnosis. It is found the prior distribution of different accident types in model training has little influence on the training performance, indicating the good applicability of the new approach.

Esimation of environmental source term of radioactive liquid stroage tank leakage accident in reprocessing facilities

FENG Zongyang, YANG Yapeng, WANG Renze, JIA Linsheng, WANG Ning, LIANG Boning

RADIATION PROTECTION. 2021, 41(S1): 64-69.

Abstract ( 82 )

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Radioactive liquid leakage accident is a typical radioactive accident in reprocessing facilities, which usually occurs in the equipment room. The generation of airborne radionuclides includes two processes: one is the release of inert gas from the solution during the leakage of radioactive liquid, and the aerosol generated by the interaction with air and floor; the other is the evaporation process after leakage (including before washing, before dilution and after dilution). The aerosol will deposit after being generated in the equipment room, and will be discharged to the environment after filtration along with the exhaust system of the equipment room. In this paper, a source term estimation method related to radioactive solution storage tank breach leakage accident is proposed. The estimation of accident leakage quality, leakage activity, airborne radioactivity concentration and integral concentration in equipment room, and environmental release source term are realized, which provides data support for emergency decision-making and response action.

Analysis of radiation effects between small modular reactor and large water reactor on the same site in the accident conditions

DONG Li, LIU Xinjian, CHENG Youying, LIU Senlin

RADIATION PROTECTION. 2021, 41(S1): 70-74.

Abstract ( 74 )

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Based on the actual site environmental characteristics of the small reactor in Hainan, this paper uses the more realistic puff model CALPUFF to analyze the radiation effects of the small reactor and the large reactor on the public outside the site according to the preliminary level-2 PSA source term, and the exposure characteristics of different accidents to the surrounding residents and workers are compared. According to the dose acceptance criteria for small reactors, the determination of the distances which meet the criteria under various weather conditions will help to get a deeper understanding of the accident characteristics of small reactors and the division of emergency planning zones, and provide references for related engineering practices and emergency supervision work.

Application experience of operational intervention level

WANG Jing, JIA Jinlei, LIN Xiujing, ZHU Jifeng

RADIATION PROTECTION. 2021, 41(S1): 75-78.

Abstract ( 75 )

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The design of nuclear emergency decision support system based on operational intervention level(OIL)is proposed in this paper according to the concept, application and latest development of operational intervention level. It mainly includes visual management of radiation monitoring data with operational intervention level as multi-level threshold, protection action decision-making based on operational intervention level, and monitoring plan formulation. In order to realize the above functions, the system should also construct operation intervention level database, radioactive background database and other basic resource databases.

A brief introduction of the development of standards for the safe transport of radioactive material in IAEA

ZHAN Yulin

RADIATION PROTECTION. 2021, 41(S1): 79-83.

Abstract ( 82 )

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There are five safety standard committees in IAEA, responsible for establishing or revising standards. TRANSSC is one among these committees, which is consisted of expert representatives from member states of the IAEA. This essay gives a brief introduction of the process of establishing and revising of safety standards in TRANSSC, and also reviewing process of SSR-6 and other relevant standards recently. Some suggestions are offered on facilitating Chinese experts attending the meetings of TRANSSC and to get better understanding of international practice and making contributions in the safe transport fields in IAEA.

Research on nuclear emergency system of spent fuel transportation with highway-sea-rail combined mode

SU Jianwen, LIU Kai, PAN Yongjun, LIN Xiujing

RADIATION PROTECTION. 2021, 41(S1): 84-90.

Abstract ( 68 )

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As the best solution to the long-distance transportation of large quantities of spent fuel, highway-sea-rail transport is a relatively common mode of transport in the international community. Thus, there is an urgent need to address the issue of spent fuel highway-sea-rail combined transport and the characteristics of its nuclear emergency management in China, and explores the ideas for the follow-up nuclear emergency work.In this paper, the relevant nuclear emergency laws and regulations and standards on the combined transport of spent fuel by highway-sea-rail at home and abroad were reviewed, and suggestions on the construction of China’s nuclear emergency response system for highway-sea-rail combined transportation of spent fuel were made with reference to the practice of spent fuel transportation abroad.

Enlightenment of typically foreign spent fuel trans-shipment ports technology to the construction of multimodal transport system of spent fuel in China

ZHENG Yu, LIU Yiqing, GUI Zhong, ZHAO Xiaodong, ZHANG Jianxin, ZHONG Maihao, YANG Ming

RADIATION PROTECTION. 2021, 41(S1): 91-95.

Abstract ( 70 )

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Single highway transportation system cannot meet the requirement of spent fuel transportation of the coastal nuclear power plant in China as the development of nuclear power. The multimodal transportation system including maritime transportation is the future direction of spent fuel transportation. In this paper, the technical conditions of typical spent fuel trans-shipment ports abroad are investigated, and the issues of our spent fuel shipping are analyzed. The results show that the typical foreign spent fuel trans-shipment ports are equipped with fixed lifting equipment that meets the requirements, and the ports are under closed management during the operation of ship in the ports, and the dose of exposure of the ports operators is monitored and controlled. As we have no experience of spent fuel shipping, let alone the relevant technical specifications of trans-shipment ports, this paper puts forward the experience of spent fuel shipping in foreign countries such as France, Britain and Japan. This paper also puts forward some suggestions on the existing problems of spent fuel shipping in China.

Preliminary study on the safety and legal issues of floating nuclear power plants during international transport

ZOU Xumao, WAN Lei, MA Xiaoya, GUO Jianlin, YANG Jue, CUI Jun

RADIATION PROTECTION. 2021, 41(S1): 96-101.

Abstract ( 77 )

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Floating nuclear power plants have been paid much attention around the world due to the advantages of modularization, multipurpose, flexible operation, transshipment, strong adaptability. Therefore, floating nuclear power plants could result in widespread application foreground and broad development space. However, because of the special deployment scenes and transportable characteristic in marine environment, the transportation issues of floating nuclear power plants as well as corresponding safety and legal problems would occur inevitably. This paper studies the different deployment scenes of a floating nuclear power plant, and the safety and legal issues under different deployment scenes during international transport are summarized and discussed. What’s more, the possible suggested solutions for international transport from the legal, security, and emergency aspects are analyzed emphatically. This study may produce significant meanings to the development of the floating nuclear power plants and to the making and improvement for related conventions and rules on the international transport of floating nuclear power plants.

Discussion on safety design method for preventing brittle fracture of shipping packages for radioactive material

ZHANG Xuesheng, GU Jianfeng, CHANG Liangming, ZHAN Lechang, ZHANG Yongxin

RADIATION PROTECTION. 2021, 41(S1): 102-106.

Abstract ( 56 )

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This paper introduces the safety design of shipping packages against brittle fracture based on fracture mechanics, and focuses on the calculation method of stress intensity factor in the condition that prevents from brittle fracture. Through analysis, the following conclusions can be drawn: the influence of nonlinear stress distribution and plastic zone are considered in the calculation of stress intensity factor in Appendix ZG of RCC-M Volume I,while comprehensive factors are considered. Therefore, the method is recommended.

Challenges arisen from the transportation of SMR modules

WANG Pengyi, LI Guoqiang, SUN Hongchao, ZHUANG Dajie

RADIATION PROTECTION. 2021, 41(S1): 107-112.

Abstract ( 85 )

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Due to the long build times and high construction cost of large nuclear power plant projects, small modular reactors have become a hot topic of nuclear industry around the world because SMR can resolve these issues and can be used for various applications. Despite the relatively great amount of literatures published on the safety of SMR cores, there are extremely limited information about the transportation of SMR. This paper investigates literatures and news about SMRs, summaries the challenges arisen from the transportation of SMR as follows: the transportation of large and heavy modules, nuclear fuels and sealed cores loaded with nuclear fuel, and proposes some actions to address them.

Discussion on technical issues related to safety transport of uranium hexafluoride

PAN Yuting, CAO Fangfang, LU Hong, LI Duohong, HONG Zhe

RADIATION PROTECTION. 2021, 41(S1): 113-116.

Abstract ( 101 )

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The safety requirements for uranium hexafluoride packaging and shipment are given, which reviewed in Regulations for the Safe Transport of Radioactive Material (IAEA SSR6) and Nuclear energy—Packaging of uranium hexafluoride (UF₆) for transport (ISO 7195). Based on the review of the Nuclear and Radiation Safety Analysis report of the uranium hexafluoride transport, The problem existing in the practice of uranium hexafluoride transport in China is discussed, and some suggestions are given.

International experience and key technology enlightenment of nuclear material accounting and control in spent fuel reprocessing plant

CHU Quanli, ZHANG Liang, LI Duohong, ZHANG Tianbao, TIAN Chuan, HE Jialin, WU Zhaohui

RADIATION PROTECTION. 2021, 41(S1): 117-121.

Abstract ( 89 )

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This paper investigates and analyses the situation of nuclear material control in spent fuel reprocessing plants in China, the key technology and experience of accounting and control measures for nuclear materials in commercial spent fuel reprocessing plants abroad, including the setting of material balance area and key measurement points for inventory in typical commercial spent fuel reprocessing plants, the overall design requirements and reality of accounting and control measures for nuclear materials. The concept of real-time accounting, etc. Based on the investigation results and analysis, and in view of the current situation of nuclear material control in China, some preliminary suggestions for the preparation of nuclear material control technology in commercial spent fuel reprocessing plants in China are put forward.

Study on the methodology of the establishment of FEPs list for the near- surface disposal of low and medium level radioactive waste in China

LI Yang, LUO Kai, CHEN Yunli

RADIATION PROTECTION. 2021, 41(S1): 122-125.

Abstract ( 92 )

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It is one of the key processes to identify all the factors affecting the long-term security of the disposal site, i.e. features, events and processes (FEPs). At present, FEPs identification and scene development are not carried out in the environmental impact assessment of low and medium level solid waste disposal site in China. Taking a disposal site as an example, this paper discusses how to establish FEPs list of low and medium level radioactive waste disposal in China.

Overview of radiation shielding design for medical proton and heavy ion accelerator

YANG Bo, SU Youwu, YAN Weiwei, LI Wuyuan, LI Yang, MA Fuhong, LI Zongqiang, MAO Wang, WANG Lijun, LIU Xuebo, LUO Changli, LI Jingfeng

RADIATION PROTECTION. 2021, 41(S1): 126-132.

Abstract ( 188 )

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In recent years, the cancer treatment industry of medical proton/heavy ion accelerator in China has started to develop rapidly. Secondary radiation is produced when the accelerator is in operation, which endangers the radiation safety of the environment, the public and workers. Reliable radiation shielding design is a necessary guarantee for the radiation safety during the operation of the accelerator. In this paper, the general considerations for radiation shielding design of medical proton/heavy ion accelerator are briefly analyzed, and several common methods of shielding calculation are introduced with an example of calculation. The research in this paper has certain reference significance for the radiation shielding design of medical proton/heavy ion accelerator to be built in the future.

Establishment of X-ray air kerma (diagnostic level) standard device

LI Yin, WEI Yingjing, CHEN Shuangqiang, FANG Dengfu, CUI Wei, FENG Mei

RADIATION PROTECTION. 2021, 41(S1): 133-138.

Abstract ( 152 )

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In order to achieve the accurate measurement of diagnostic level X-ray dose rate value, and to carry out the traceability and transmission of it, this article established a tube voltage of 40-150 kV and dose rate of 1.0×10^-3-10 Gy/h X-ray air kernal energy (diagnostic level) standard device according to the requirements IEC 61267—2005 standard, experimental measurement gives out the radiation field characteristics under different radiation qualities, and relative expanded uncertainty of air kernal energy rate in radiation field getting by evaluation. Experimental measurement results show that the radiation quality of RQR 5,while in a distance of 1.0 m from focal point of the X-ray spot, uniformity of radiation field >99.0% is ø8.0 cm, and uniformity >95.0% is ø10.0 cm; the contribution of scattering to radiation field is <0.9%; it is less than 2.0% about inverse square of the distance in the range of 1.0-5.0 m; the repeatability of radiation dose rate measured by the standard ionization chamber is 0.1%. Radiation qualities with RQR2-RQR10 series and RQT8-RQT10 are measured in HVL and homogeneity coefficient, which HVL deviation does not exceed ±0.09 mm, and homogeneity coefficient deviation does not exceed ±0.02. The evaluation result of relative expanded uncertainty in radiation field air kernal energy rate is U_rel=3.8%(k=2). Experimental measurement results of radiation quality characteristics show that the performance indicators of X-ray air kerma (diagnostic level) standard device meet the requirements of standards such as IEC 61267—2005.

Establishment of X-ray air kerma (radiotherapy level) standard device

CHEN Shuangqiang, WEI Yingjing, LI Yin, FANG Dengfu, TANG Zhihui, YANG Bo

RADIATION PROTECTION. 2021, 41(S1): 139-144.

Abstract ( 98 )

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In order to carry out the work of traceability and transmission of X-rays at the dose rate of radiotherapy, according to IEC 60731—1997 standard, a standard device of X-rays air kerma (radiotherapy) with tube voltage of 10-250 kV and dose rate range of 1.0×10^-3-10 Gy/h was established. Of which 10-60 kV X-ray air kerma (radiotherapy) standard device at 1.0 m heterogeneity is less than 1% of the radiation field for ø60 mm, scattering contribution to radiation field is less than 1.2%, from the source 1 m-5 m within the inverse square law at 2.5%, within the scope of using standard ionization chamber stability of the measurement device is 1.8%, repeatability is 0.1%. 60-250 kV X-ray air kerma (radiotherapy) standard device at 1.0 m heterogeneity is less than 1% of the radiation field for ø80 mm, scattering contribution to radiation field is less than 1.2%, from the source 1 m~5 m within the inverse square law at 1.5%, within the scope of using standard ionization chamber stability of the measurement device is 1.7%, repeatability is 0.03%. The relative extended uncertainty of the radiation field air kerma (rare) of the standard device was 3.0% (k=2). The performance indexes of the measured device all met the verification/calibration requirements of the therapeutic level dose detection instrument.

Establishment of γ-ray air kerma (radiotheraph lever) standard device

LI Weiming, WEI Yingjing, LI Yin, CHEN Shuangqiang, WANG Mingliang, HAO Shidong

RADIATION PROTECTION. 2021, 41(S1): 145-150.

Abstract ( 89 )

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A standard device for γ-ray air kerma (radiotherapy) was established using ⁶⁰Co radioactive source with an activity of 1.85×10¹⁴Bq. Meanwhile, a safety interlock system with six subsystems and logic of “and” gate was designed for the standard device. In order to evaluate the performance indicators of the standard device, the air kerma rate range of the radiation field, the radiation field, uniformity and scattering were measured in sequence. The measurement results show that at a distance of 1 m to 5 m, the air kerma rate of the radiation field without lead attenuation ranges from 1.71 Gy/h to 43.5 Gy/h, and the minimum air kerma is after adding the attenuator The kerma rate can reach 1.5×10^-4 Gy/h. Under the condition of no attenuator, the inverse square law of the standard device is established within 1% in the range of 1 m to 5 m from the radiation source. Under the condition of adding the attenuator, the inverse square law of the standard device is established within 3.5% in the range of 1 m to 5 m from the radiation source. At a position of 2 m, the radius of the radiation field where the air kerma rate fluctuates within 5% is 13 cm, and the radius of the radiation field where the air kernal energy rate fluctuates within 1% is 7 cm. The technical performance indicators of the standard device meet the verification/calibration requirements of the radiotherapy dosimeter.

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