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Objective To provide a theoretical basis for radiation health supervision through an analysis of the situation of computed tomography (CT) equipment quality control and CT room radiological protection in Guangdong Province, China in recent years. Methods We collected the data of 392 times of CT quality control and radiological protection testing by a third-party radiological health technical service institution in Guangdong Province from 2019 to 2021. We analyzed the levels of CT-owning hospitals, CT manufacturers, CT quality control test results, and the pass rate of radiation protection tests. Results The examined CT scanners were from different levels of hospitals in Guangdong Province, and were manufactured by nine major CT equipment manufacturers at home and abroad. The pass rate of CT room radiological protection was 99.88%, and the ambient dose equivalent rates of five monitoring points exceeded the limit, with four at the control room door and one at the shield wall of the room. The overall pass rate of CT equipment quality control was 99.49%, and the non-conforming parameters were the accuracy of positioning light and the deviation of reconstructed slice thickness. Conclusion In recent years, CT equipment quality control and room radiation protection in Guangdong Province have been at a high level.
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Introduction Anesthetists play an important role during interventional radiology procedures. Like the main operator, anesthetists may also be subject to significant radiation levels in the fluoroscopy suite. Due to its complexity, hepatic chemoembolization procedures demand high fluoroscopic times and digital subtraction angiography images, exposing patients and medical staff to high radiation doses. Objective To assess and quantify the radiation to which one anesthetist was exposed over the course of seven consecutive hepatic chemoembolization procedures, and compare it to the exposure received by the main operator. Methods Medical staff dosimetry was evaluated during seven consecutive hepatic chemoembolization procedures conducted in a private hospital in Recife (Brazil), using thermoluminiscent dosimeters placed in regions of the head and torso. Results For the seven procedures evaluated in this study, the anesthetist received, on average, absorbed doses to the glabella, left eyebrow, right eyebrow and effective dose of 142.4 ± 72 µSv, 117.3 ± 66 µSv, 137.8 ± 71 µSv and 12.4 ± 8.4 µSv, respectively. Conclusions In some cases, ocular dose and effective dose received by the anesthetist may be 4 and 4.7 times greater, respectively, when compared to the main operator. According to the results of this study, the current occupational annual dose limit to the lens of the eye of 20 mSv can be exceeded with only two hepatic chemoembolization procedures per week if adequate radiation protection conditions are not guaranteed.
Introducción El anestesiólogo desempeña un papel importante durante los procedimientos de radiología intervencionista. Al igual que el operador principal, el anestesiólogo también puede estar expuesto a niveles significativos de radiación en la sala de fluoroscopía. Debido a su complejidad, durante los procedimientos de quimioembolización hepática se deben utilizar imágenes de fluoroscopía y angiografía de sustracción digital por períodos prolongados, exponiendo a los pacientes y al personal médico a dosis elevadas de radiación. Objetivo Evaluar y cuantificar la radiación a la cual se expuso un anestesiólogo durante el transcurso de siete procedimientos consecutivos de quimioembolización hepática, y comparar con la exposición recibida por el operador principal. Métodos Evaluación de la dosimetría ocupacional durante siete procedimientos consecutivos de quimioembolización hepática realizados en un hospital privado de Recife (Brasil) por medio de dosímetros termoluminiscentes ubicados en regiones de la cabeza y el torso. Resultados Para los siete procedimientos evaluados en este estudio, el anestesiólogo recibió, en promedio, dosis absorbidas en el entrecejo, la ceja izquierda, la ceja derecha y dosis efectivas de 142,4 ± 72 µSv, 117,3 ± 66 µSv, 137,8 ± 71 µSv y 12,4 ± 8,4 µSv, respectivamente. Conclusiones En algunos casos, la dosis ocular y la dosis efectiva que recibe el anestesiólogo puede ser, respectivamente, entre 4 y 4,7 veces más alta que la que recibe el operador principal. De acuerdo con los resultados de este estudio, el límite ocupacional anual de dosis en cristalino (20 mSv) se puede superar con apenas dos procedimientos de quimioembolización hepática a la semana en caso de no garantizarse las condiciones adecuadas de protección contra la radiación.
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Objective:To study the actual effect of the use of personal protective equipment of the examined individuals, and provide reference and basis for the correct use of personal protective equipment and the radiological health administrative law enforcement.Methods:From February to June 2022, the imaging department of Qingdao Municipal Hospital selected 170 patients who underwent X-ray imaging examination (oral panoramic, dental radiography, DR photography, CT scanning), including 25 with oral panoramic and dental radiography, 60 with CT scanning and 60 with DR imaging. The thermoluminescent dosimeter was used to detect the ambient dose equivalent at the point of concern for 170 examined individuals who have used personal protective equipment to cover their sensitive parts, and to analyze the data detected at the same point as above when routinely using the same equipment.Results:There was a statistically significant difference in the dose equivalent at the same points inside and outside the lead neckband ( t=-2.23, P<0.05). There was no statistically significant difference in the dose equivalent at the same point inside and outside the lead collar during dental radiography ( P>0.05). During DR photography (chest PA, lateral and lumbar AP), the examined individuals were wearing lead aprons. Among them, there was a statistically significant difference in the dose equivalent at the same points inside and outside the lead aprons of children′s chest PA and adults′ chest PA ( U=10.00, 19.00, P<0.05). There was no statistically significant difference in the dose equivalent at the same points inside and outside the lead aprons of adult′s chest PA and lumbar AP ( P>0.05). When performing CT scan (chest or upper abdomen), there was a statistically significant difference in the dose equivalent at the same points of wrapped lead aprons( U=878.50, 11.00, P<0.05). Conclusions:The correct use of personal protective equipment is a complex technical problem. It is very important to fully and accurately understand the optimization principle of radiation protection and correctly use personal protective equipment for the examined individuals. The administrative punishment of radiation health on the use of personal protective equipment of the examined individuals should be cautious.
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Objective:To explore the present state of radiation protection in an operational large-sized irradiation facility and the distribution of incurred radiation doses in the controlled area under abnormal conditions, and to analyze the acompanied radiation risk.Methods:With an irradiation facility in operation as the research object, the radiation doses were measured using AT1121 X and gamma dose rate meters for the soruce both in working and storage. Thermoluminescent dosimeters (TLDs) were used to measure the dose distribution in an irradiation field under abnormal conditions. The radiation safety features were checked against the several relevant national standards, with the radiation risks evaluated.Results:The radiation dose rates were in the range of 0.09-0.11 μSv/h, lower than the environmental background level whether in storage or working. Under abnormal working conditions, the radiation dose values from high to low were 1.0-101.3 Sv, 32.7-514.0 mSv and 8.7-183.2 μSv in the irradiation field, respectively. At the outside maze, the doses were close to the background level.Conclusions:The protection features of the irradiation facility meet the requirements of the relevant national standards. Under abnormal conditions, radiation could cause serious damages to the persons staying in the irradiation field. These persons were suggested to access to maze as soon as possible to reduce the exposure time, and activate the emergency protection equipments to deescalate the 60Co source onto the well.
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Objective@#To investigate the changes of ambient dose equivalent rate in 99mTcO4- single photon emission computed tomography (SPECT) of the thyroid among patients with hyperthyroidism, so as to provide insights into radiation protection guidance.@*Methods@#Patients with hyperthyroidism who underwent 99mTcO4- SPECT of the thyroid in a tertiary hospital were enrolled. The ambient dose equivalent rate was measured at different time points following 99mTcO4- infection and at sites with different distances from patients' neck, and the effects of time post-injection, distance from patients' neck, 24-hour thyroidal radioiodine uptake and thyroid weight on the ambient dose equivalent rate were examined using a generalized linear mixed model.@*Results@#Totally 100 patients with hyperthyroidism were enrolled, including 24 men and 76 women and with a mean age of (38.5±14.0) years. The generalized linear mixed model was statistically significant (F=6 610.165, P<0.001), and patients' thyroid weight, time post-injection and distance from patients' neck significantly affected the ambient dose equivalent rate (F=57.967, 15 988.574, 11 200.645, all P<0.001), and the ambient dose equivalent rate positively correlated with patients' thyroid weight and negatively correlated with time post-injection and distance from patients' neck.@*Conclusions@#The ambient dose equivalent rate is affected by patients' thyroid weight, time post-injection and distance from patients' neck among patients with hyperthyroidism undergoing 99mTcO4- SPECT of the thyroid. Delay in contact with patients or keeping distance from patients may be effective for radiation protection.
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Background Individual monitoring of occupational external exposure is an essential part of the occupational health management of radiation workers, and is an important basis for the evaluation of individual absorbed dose and the diagnosis of occupational radiation diseases. Continuous participation of monitoring service providers in intercomparison is a fundamental quality assurance for routine monitoring, which can identify problems and improve them in time. Objective Taking the Laboratory of Radiation Protection in Shanghai Institute of Preventive Medicine as an example, to evaluate the performance of an individual occupational external dose monitoring system in the laboratory, identify influencing factors of the monitoring results, and provide a basis for improving the quality of daily monitoring by analyzing the process and results of a national intercomparison of individual dose monitoring. Methods According to the Testing criteria of personnel dosimetry performance for external exposure (GBZ 207-2016), and the relevant requirements of Class II (photon) inspection, a total of 20 groups of blind sample dosimeters were measured for four consecutive years from 2018 to 2021. The radiation energy source of each group was identified, and related personal dose equivalent Hp(10), the uncertainty of measurement results, and the deviation between the reported value and the reference value were calculated. The national intercomparison process and results of individual dose monitoring were also analyzed. Results The energy sources of the blind samples in the tested laboratory for four years were N100 or Cs-137. The reported dose values of the blind samples were 0.57-4.61 mSv, the combined uncertainties were 0.043-0.365 mSv, the expanded uncertainties (k=2) were 0.09-0.73 mSv, and the relative expanded uncertainties (k=2) were 13.8%-16.4%. The single-group performance ∣Pi∣ of 20 sets of blind samples in the four years was ≤0.10, the yearly comprehensive performance of 5 sets of blind samples was ≤0.10, and the yearly Q score of the test report was >15 points. The laboratory achieved excellent results in the national intercomparison of individual dose monitoring in four consecutive years, except the Q value not reaching full score. Conclusion The laboratory exhibits standardized data processing of individual dose monitoring, generates accurate and reliable results, and meets the requirements of relevant national standards; but it should continue to participate in the national intercomparison of individual dose monitoring, strengthen the angular response research of energy identified dosimeter, improve the monitoring ability of low-dose X-rays, analyze the key points of reducing the uncertainty of measurement results, and continuously improve the monitoring ability.
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Objective To develop a portable nuclear radiation detector with low-energy γ-nuclide recognition capability for rapid measurement of the dose levels in low-energy radiation fields and identification of nuclides. Methods A digital multi-channel circuit was developed for a detector based on the room temperature semiconductor cadmium zinc telluride, nuclide recognition was achieved using an intelligent nuclide recognition algorithm, and the energy response function G(E) was used to calculate the real-time ambient dose equivalent rate H*(10). Results The portable spectrometer had a minimum detectable energy of 20 keV, and the typical energy resolution for low-energy X-rays was > 4.10% at 59.5 keV and 20℃, enabling accurate identification of 241Am nuclide. Conclusion The device has a good measurement performance for low-energy γ/X rays, effectively addressing the limitations of existing devices for monitoring low-energy radiation fields, and provide reliable technical methods for monitoring and emergency response in spent fuel reprocessing plants or nuclear material production plants.
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Objective To study the ambient radiation of patients undergoing 18F-FDG PET/CT myocardial metabolism imaging, and to provide data for protection of surrounding people against radiation. Methods A total of 33 patients were selected for 18F-FDG PET/CT myocardial metabolism imaging. Dose equivalent rate was measured at the height of patient chest, in different directions, at different time points and at different distances, to investigate the distribution of ambient radiation of the patient. Results At the same time point and distance, the dose equivalent rates at the left and right sides of the patient were lower than the front and back sides. The dose equivalent rate at 1 m in front of the patient was 13-21 μSv/h after 18F-FDG injection, and decreased to 5-14 μSv/h after PET/CT imaging, with a mean decrease of 46%. The ambient dose equivalent rate decreased exponentially with distance (10~300 cm), and the mean power was −1.2. Conclusion The ambient radiation of patients undergoing 18F-FDG PET/CT myocardial metabolism imaging was high after 18F-FDG injection, and the ambient dose equivalent rate decreased rapidly with time and distance. Our results suggest that patients undergoing myocardial metabolism imaging should avoid prolonged and close contact with other people on the day of examination.
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Objective To detect the radiation of 131I in treatment site of a grade A tertiary hospital. Methods A total of 25 patients with thyroid cancer were administrated 131I at a total dose of 82880 MBq. After administration, the ambient dose equivalent rate of the ward was detected with X- and γ-ray detectors. After patient discharge, surface contamination of the ward was detected with α/β surface contamination meter. During patient hospitalization and on the day of discharge, air samples were collected from 131I treatment site and office area. The air samples were measured using a HPGe γ-ray spectrometer and the concentration of 131I in air was calculated. Results The ambient dose equivalent rate in the ward ranged from 0.15 to 0.46 μSv/h. Before ward cleaning, surface contamination ranged from 0.53 to 40.1 Bq/cm2 and the highest value was recorded on the toilet. Within 4 h after administration, the concentrations of 131I in air in treatment site and the corridor of the office area were 1.74 Bq/m3 and 0.66 Bq/m3, respectively. The ventilation air flow rate in the treatment site was 0.50 m/s. Ventilation decreased the concentration of 131I in air by 29.7%, 79.7%, and 53.3% compared with the previous day during hospitalization and on the day of discharge. Conclusion The radiation of external exposure of 131I in the treatment site is low and the shielding is effective. Before ward cleaning, the surface contamination is lower than the required limits except for the toilet. Ventilation is the primary way to reduce the concentration of 131I in air.
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Objective To investigate the radiation dose to operators in the process of 241Am-Be neutron source logging, and discuss neutron source management and protective measures for operators in well logging. Methods Through on-site observation and measurement of 241Am-Be neutron source logging in a company, we obtained the surface γ dose rate and neutron dose rate of the neutron source, as well as the operating time and distance of various processes including source taking, transfer, and loading, calculated the radiation dose to operators in various processes, and analyzed the source and proportion of the personal effective dose to operators. Results The effective doses of neutron irradiation and γ irradiation were 94.17 μSv and 2.72 μSv, respectively, for the combined processes of source tank inspection, transfer, and detection; 36.66 μSv and 24.08 μSv, respectively, for source loading and unloading; and 130.83 μSv and 26.80 μSv, respectively, for the whole neutron source logging process. The total annual effective dose of neutron source logging was 15.78 mSv, as estimated by logging 100 times per year. Conclusion In the process of 241Am-Be neutron source logging in the company, the effective dose to operators mainly arises from neutron irradiation. Therefore, it is necessary to strengthen neutron source management and take effective protective measures against neutron radiation.
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Objective:To study the variation in activity in patient′s body with differentiated thyroid cancer (DTC) treated with 131I and external dose level, analyze the relationship between the both, and estimate the correction factor for the dose equivalent rate for the patients with residual activity of 400 MBq. Methods:A total of 43 DTC patients who received 131I therapy for the first time after total thyroidectomy were studied. The dose was 1 850-3 700 MBq and average dose was (2 405±777) MBq. The measurements of residual activity in patient′s body and of dose equivalent rate at 0.3, 1 and 3 m in front of the patients were performed at 2, 6, 20, 22, 24, 27, 30, 44, 46, 48, 54, 68 and 72 h after administration of 131I. Results:The residual activity in patient′s body after 131I therapy varied with time as a function of A= A0 (1.033 16e -0.062 4t+ 0.017 17). It can be estimated that the effective half-life of DTC patients treated with thyroid remnant 131I ablation therapy is 12.19 h. It needs only 26.4-38.9 h to reduce the internal activity to the 400 MBq. The functions of variation with time of normalized dose equivalent rate at 0.3, 1, and 3 m away from patients were: H· 0.3=127.220 7e -0.054 8t+ 3.765 71; H· 1=30.225 8e -0.064 4t+ 0.824 67; and H· 3=4.161 9e -0.061 5t+ 0.167 97, respectively. There was a positive correlation between residual activity and dose equivalent rate at 1 m ( r=0.982, P<0.05), and the function is H· 1=0.025 A+ 1.245. When residual activities in DTC patient′s body were 1 000, 700 and 400 MBq, the corresponding dose equivalent rates at 1 m from patients were 26.2, 18.7 and 11.2 μSv/h, respectively. The correction factors for dose equivalent rate at 0.3, 1 and 3 m from patients with 400 MBq were 0.25, 0.49 and 0.70, respectively. Conclusions:DTC patients with administration of 131I activity below 3 700 MBq need only to be hospitalized for two days to reach the discharge standards. When the residual activity in DTC patient′s body drops to 400 MBq, the dose equivalent rate at 1 m is far less than 25 μSv/h. Simply using the point source formula to estimate the dose equivalent rate around the patient will result in overestimation. Therefore, the correction factor used in the estimation of radiation doses to patients by using the formula needs to be further studied so as to make the model-based estimated result more consistent with the actual situation.
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Objective:To compare the calculation result and analyzes the reasons for their differences so as to provide reference for the revision and improvement of the current national standards on radiation shielding design for the room of brachytherapy.Methods:For the initial activity 10 Ci (1 Ci=3.7×10 10 Bq) of radioactive sources, the shielding schemes of brachytherapy room were designed in accordance with UK Institnte of Physics and Engineering in Medicine(IPEM) Report 75, USA NCRP Report 151 and the national standard GBZ/T 201.3-2014, respectively. The differences in shielding limits, occupancy factors and other relevant factors are compared in detail. Results:The annual exposure time in a typical brachytherpy room was about 330 h. The point-specific concrete thickness were 70, 65, 61, 70, 50 cm as required by NCRP Report 151, 41, 43, 30, 40, 39 cm by IREM regulations and 84, 79, 46, 88, 39 cm by GBZ/T 201.3, respectively. The concerned concrete shielding thickness calculated under the GBZ/T 201.3-2014 was generally thicker, with lesser difference from NCRP Report 151 result, whereas that from the IPEM75 report was thinnest. The equivalent lead shielding thicknesses of the protective doors calculated using the three method are 1.170, 0.854 and 1.040 cm, respectively.Conclusions:The shielding thickness calculated using the calculation method and evaluation index recommended by the current Chinese shielding standards for brachytherapy bunker is similar to that reported in NCRP151, but is conservative. In particular, the evaluation index of instantaneous dose equivalent rate required by the current national standards and the relative conservative value of occupancy factor will significantly increase the shielding thickness required by the main shielding area.
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Objective:To explore the existing issues in radiation protection during the treatment of 131I by means of measuring the ambient dose equivalent rate to patients with thyroid cancer and the dose equivalent to the surface of chest of patients during hospitalization. Methods:The ambient dose equivalent rate (peer) was measured by using gamma ray detector for selected 78 patients who received 131I treatment in a hospital 10 min, 1 d, 2 d, 3 d and 5 d after administration with 131I. The measurements were made at distances of 5 cm, 0.5 m and 1 m from the body surface in front, rear, left and right directions. The photoluminescence dosimeter on the chest of the patients was used to measure the effective dose during hospitalization period (6 d). Results:The ambient dose equivalent rate on the surface of chest of patients was up to 4.81 mSv/h 10 min after administration of medicine. The dose equivalent on the surface of chest of patients before discharge ranged 2.6-64.1 μSv/h. The cumulative dose on chest surface during hospitalization was 15.9-58.8 mGy. There was a significant difference in the dose rate at 5 cm from the body surface between 3.7 GBq group and 5.55 GBq group 10 min after medication ( t=-6.11, P<0.05). There was a significant difference in the dose rate at 5 cm from the body surface between male and female groups 10 min after medication ( t=4.52, P < 0.05). There was no significant difference in other groups ( P > 0.05). Conclusions:During the 131I treatment, patients had high level of radiation around them, so it is necessary to strengthen the protection and management of patients and reduce unnecessary exposure to the public.
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Objective To investigate the effect of iron shield at different depths within main protection wall on the dose rate outside the protection wall. Methods By adopting the FLUKA code, a therapeutic room model was constructed with its primary protective barrier consisting of concrete and iron. In order to obtain its ambient dose equivalent rate distribution, the 250 MeV protons and 220 MeV protons impinging on water phantom were simulated separately. Results With varying depth of iron plate embedded in barrier, the ambient dose equivalent rates in the two simulated conditions differed sinificantly at 30 cm outside the protection wall. The maximum ambient dose equivalent rate(220 MeV:3.42 μSv/h, 250 MeV:6. 39 μSv/h) was more than 2 times higher than the minimum ambient dose equivalent rate ( 220 MeV:1. 75 μSv/h, 250 MeV: 3. 32 μSv/h ) . Conclusions In the design of therapeutic proton accelerator, it is essential to evaluate carefully the location where the iron shield is in main protection wall.
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Objective@#To investigate the effect of iron shield at different depths within main protection wall on the dose rate outside the protection wall.@*Methods@#By adopting the FLUKA code, a therapeutic room model was constructed with its primary protective barrier consisting of concrete and iron. In order to obtain its ambient dose equivalent rate distribution, the 250 MeV protons and 220 MeV protons impinging on water phantom were simulated separately.@*Results@#With varying depth of iron plate embedded in barrier, the ambient dose equivalent rates in the two simulated conditions differed sinificantly at 30 cm outside the protection wall. The maximum ambient dose equivalent rate(220 MeV: 3.42 μSv/h, 250 MeV: 6.39 μSv/h) was more than 2 times higher than the minimum ambient dose equivalent rate(220 MeV: 1.75 μSv/h, 250 MeV: 3.32 μSv/h).@*Conclusions@#In the design of therapeutic proton accelerator, it is essential to evaluate carefully the location where the iron shield is in main protection wall.
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Objective To estimate the hand dose of occupational staff in five procedures of interventional radiology.Methods The thermoluminescence ring dosimeter(TLD,LiF:Mg,Ti)calibrated in dose equivalent Hp(0.07) was used to monitor the radiation dose to the both hands in five procedures of interventional radiology in four hospitals.Meanwhile,the tube voltage and current and the fluoroscopy time,accumulated doses,DAPs (Dosed-Area Product) and photographic frames were also recorded for statistical analysis(SPSS 18.0).Results In this study,five interventional procedures were monitored in 119 cases.The doses to the left and right hands of occupational personnel in the five interventional procedures were analyzed,and the difference was statistically significant (t =1.99,P＜ 0.05).The differences in the dose to the left hand and the right hand of the first operator with different interventional procedures were statistically significant (F =455.83,116.45,P＜0.01).Tube voltage,current,fluoroscopy time,and photographic frames in influence factor analysis were statistically significant (r =0.570,0.712,0.564,0.711,P＜ 0.05),indicating that four factors might affect the operator's hand doses.In addition,increasing these four factors directly led to increase in the operator's hand doses.The variables with statistical significance in the above single factor analysis were introduced into the multiple linear regression equation and the stepwise regression method was used to fit the equation.The fitting equation was y=225.763+ 1.862x1-98.125x2 (F=22.726,P＜0.05).Where,x1 was the fluoroscopy time and x2 was the photographic frames which showed that they were the main factors affecting the hand dose.Conclusions The dose to the hands of the primary operator was the highest,followed by the second operator,the assistant or nurses.The order of average doses to the hands of the primary operator was PM＞RFA＞CA＞ PTCA+PCI ＞ ITCA in the five procedures.The annual equivalent dose to primary operator's hands may exceed the dose limit for hands in practicing large amount of the PM procedures.
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Objective To measure and analyze the neutron dose equivalent rate produced by an IORT accelerator with 9 and 12 MeV electron energyies,and compare them with those from a Siemens Primus linear accelerator with the same electron energy,in order to provide data reference for the risk of secondary cancer induced by radiotherapy.Methods Using the neutron detector LB6411,the neutron dose equivalent rates produced by the IORT accelerator of 9 and 12 MeV were measured on some key locations,such as the head of the accelerator,cylinder bottom,patient plane with electron energies 9 and 12 MeV.The similar measurements were also performed on the same locations on a Siemens conventional accelerator.The data were collected and analyzed and the result wer compared between the two accelerators.Results Neutron dose equivalent rates from the IORT accelerator with 9 MeV energy were (51.8±3.1),(45.5 ±1.5),(70.5 ±4.9) and (68.2±3.3) μ Sv/h near the head of the accelerator,cylinder bottom,patient plane,with 5.9％,5.4％,17.8％ and 21.5％ lower than at 12 MeV,respectively.The dose equivalent rates at the similar locations from the Siemens Primus accelerator were (277.3 ±1.2),(285.1 ±1.6),(185.1 ±1.8) and (182.8 ±2.4) μSv/h at 9 MeV,with 48.8％,47.6％,48.7％,52.2％ lower than those at 12 MeV,respectively.At the energy of 12MeV,the neutron equivalent dose rate from the IORT was lower by a factor of about 10 than for Siemens Primus accelerator.Conclusions The neutron dose equivalent rates generaged by both the IORT and the Siemens Primus are higher at 12 MeV than at 9 MeV,which would lead to an increased risk of secondary cancer to patients.The traditional medical accelerator produces much higher neutron dose equivalent rates than the intraoperative electron accelerator,for which the appropriate shielding should be takn.
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Objective To validate and discuss the time response correction formula for four types of dosimeters (6150AD6 + 6150AD-b,FH40G + FHZ672E-10,451P ionization chamber and AT1123).Methods The ambient dose equivalent rates shown by survey meters were recorded separately when X-ray emission time was 500,200,100 and 50 ms.The corrected values were obtained by the formula of circuit having a capacitance C and asistance R in series.Results Therewas no correlation between the value measured by AT1123 dosimeter and the time of irradiation.The values by other three kinds of dosimeters obviously varied with the time of irradiation.Conclusions It is not required to make the time response correction for the measured value of ATl123 dosemeter,whereas the values measured by the other three dosimeters could be corrected by the time response correction formula.
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Objective To evaluate the reliability of HARSHAW-3500 thermoluminescence dosimetry system by testing its performances.Methods HARSHAW-3500 thermoluminescence dosimetry system had its performances tested and evaluated according to Verification regulation of thermoluminescence dosimetry systems used in persontal and environmental monitoring forXandgammaradiation(JJG 593-2006),Testingcriteriaofpersonneldosimetryperformanceforexternalexposure (GBZ 207-2016),Specifications for individual monitoring of occupational external exposure (GBZ 128-2016) and Thermoluminescence dosimetry systems for personal and environmental monitoring (GB/T 10264-2014),such as batch homogeneity,repeatability,linearity,incidence angle response,stability,energy response and scale factor,quantity inspection,residual dose,detection limit and etc.Results Testing results of various performance indicators proved to be within the limits according to national and industrial standards.Conclusion HARSHAW-3500 thermoluminescence dosimetry system conforms to the requirements for radiation dose measurement.It is beneficial to the improvement of quality and performance of thermoluminescence dosimetry by performances analysis and evaluation.
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Objective To improve the quality of personal dose equivalent measurement by exploring optimal annealing temperature conditions.Methods Totally 60 pieces of thermoluminescent detectors were randomly and equally divided into 6 groups.The 6 groups underwent 10-min annealing under 200,220,230,240,250 or 260 ℃ respectively,and then were cooled with the same conditions and went through measurement after irradiation by the calibrated radiation source.The above operation of annealing,cooling and measurement were repeated for 10 times,and the 6 groups were compared on dispersity,sensitivity and glow curve.Results Single test proved that under 240 ℃ the dispersity,sensitivity and glow curve gained optimal results comprehensively,while repeated tests showed that the dispersity had the optimal value under 250 ℃ and the sensitivity decreased significantly as the times of annealing rose.Conclusion Annealing conditions have to be selected according to the requirements of the thermoluminescent detector.