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Objective:To evaluate 90Y activity concentration in renal excretions during the first 48 hours after being treated with 90Y resin microspheres seleceive internal radiation therapy(SIRI) and to provide advice on the management of patient excreta after surgery. Methods:After surgery, urine excreted from 3 patients during 0-24 h and 24-48 h was collected respectively, and the 90Y activity concentration in urine was tested and analyzed. Results:90Y radioctivity in the urine excreted from 3 patients after surgery was (1 266±258)kBq/GBq during 0-24 h and (140±106) kBq/GBq during 24-48 h, respectively, and 90Y activity concentration were (640±113) kBq/L during 0-24 h and (53±12) kBq/L during 24-48 h. Conclusions:90Y radioactivity in patient′s urine excreted at 1 d was about 10 times higher than that at 2 d. After surgery, patients can accelerate the reduction of free 90Y activity by increasing excretion. Urine excreted by the patients during hospitalization should be handled in accordance with the requirements of the national standard HJ 1188-2021 Radiation protection and safety requirements for nuclear medicine.
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Objective:To conduct radiation detection and dose assessment in selective internal radiotherapy with 90Y resin microspheres for the purpose of providing a reference for radiological protection. Methods:The dose rates from radiotherapy with 90Y resin microspheres were detected in the process of each operation at different distances from the body surface of patients the levels of dose to the persons concerned were compared with the relevant national regulations and standards. Results:The estimated dose rates were 1.12-454 μSv/h during 90Y resin microspheres dispensing and 2.06-58.2 μSv/h during surgical procedure. The dose rates at 0.5 h after surgery were 22.7-64.1 μSv/h at 5 cm and 0.82-2.55 μSv/h at 1 m from three patient′s body surface. Assuming treating 200 patients a year, the annual individual effective dose to the radiation workers was 0.12-1.03 mSv/year. The annual individual effective dose to the public, comforters and carers of patients was 0.02-0.24 mSv/year after release of a patient. Conclusions:During the treatment, nursing and release of patients, the radiation doses to workers, carers and the public are lower than the individual dose limit given in the GB18871-2002 basic standards for protection against ionizing radiation and for the safety of radiation sources and the management target value set by of the relevant medical institutions.
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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 verify and discuss the consistency and applicability of the semi empirical formula and Monte Carlo simulation method in the radiation shielding calculation for high energy synchrotron radiation source.Methods:The semi empirical formula and Monte Carlo simulation were used to calculate the ambient dose equivalent outside of the shielding.Results:The ratio of Jenkins semi empirical formula result to Monte Carlo simulation result was 111%-153%. The ratio of Sakano semi empirical formula result to Monte Carlo simulation result was 201%.Conclusions:For a single shielding material, the semi empirical formula can be simple and conservative to complete the shielding calculation for high-energy electron accelerator. For a variety of shielding materials, Monte Carlo simulation method should be used.
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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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To ascertain the background and frequency of diagnostic medical X-ray procedures in China and provide the basis for regulatory oversight of such applications,a total of 557 medical institutions in 25 provinces or municipalities were surveyed by means of the optimally designed questionnaires and through stratified quota sampling.The numbers of procedures were calculated in terms of the type of procedures and the sex and age of examined patients.As a result,the frequencies of diagnostic X-ray procedures for 2016 in the country were derived using multiple linear regression analysis.The frequency of X-ray diagnosis in 10 provinces of China in 2016 was estimated to be 379-1 228 examinations per 1 000 population.Diagnostic X-ray applications have shown a rapid expansion in 2016 as compared with the period of "9th Five-Year Plan".It is very important to strengthen the regulation of medical diagnostic X-ray applications.
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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 increase the cumulative measurement level of 222 Rn and 220 Rn and ensure the accuracy and reliability of the measurement result . Methods By using improved 222 Rn-220 Rn discriminative detectors ( LD-P detectors) , the radon research group of National Institute for Radiological Protection Chinese Center for Disease Control and Prevention participated with the intercomparison organized by National Institute of Radiological Science ( NIRS) , Japan. Specifically, with the 222 Rn-220 Rn discriminative detectors being sent to Japan, the comparison was completed under different conditions in the 222 Rn chamber and 220 Rn chamber in NIRS. After exposure, the detectors were sent back to our laboratory for etching and analysis, and then measurement result were informed to NIRS. Finally, NIRS returned the exposure reference values of 222 Rn and 220 Rn to our laboratory. Results Under the conditions of high and low levels of 222 Rn, the relative percentage differences ( RPD ) between the measured values and the reference value provided by the NIRS were -12. 0％ and -11. 8％, respectively, while coefficients of variation ( COV) were 3. 0％ and 6. 2％, respectively. Under the conditions of high level and low levels of 220Rn, the relative percentage differences (RPD) between the measured value and the reference value provided by the NIRS were -0. 8％ and -8. 0％, respectively; coefficients of variation ( COV ) were 6. 7％ and 4. 5％, respectively. Conclusions This intercomparison result were categorized by NIRS ( PRD<10％) , with the satisfactory result of LD-P detectors available.
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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 measure the circuit time constants of 4 kinds of radiation survey meters (451P ionization chamber dosimeter,6150AD6 + 6150AD-b dose meter,FH40G + FHZ672E-10 dose meter and AT1123 dose meter) and,to discuss the formula of time response correction and its application.Methods In the condition of continuous exposure of X-ray machine,the ambient dose equivalent rates shown by survey meters were recorded.In order to get the circuits time constant,the least squares fittingmethod was used to fit the data using the time response formula of circuit having a capacitance C and a resistance R in series.Results The relative uncertainty of fitted circuit time constants was higher than 20％ except for 6150AD6 + 6150AD-b dose meter.The relative uncertainty of fitted r was 8％ for 6150AD6 + 6150AD-b dose meter.Conclusions The time required to stabilize the dosimeter readings was 8,5,3 and 2 s,respectively,for the 451P ionization chamber dosimeter,6150AD6 +6150AD-b dose meter,FH40G + FHZ672E-10 dose meter and AT1123 dose meter.The rising trend of their measured values was not fully accordance with the RC circuit time response correction formula.
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Objective To use Monte Carlo method to build a shielding calculation model for the proton and heavy ion treatment room,and to provide a reliable calculation method for shielding design.Methods A Monte Carlo-based FLUKA code was adopted to build the shielding calculation model for the proton and heavy ion treatment room,and to simulate the radiation field distribution in the proton and heavy ion treatment room.The calculation model was verified through the radiation detection around the proton and heavy ions treatment room.Results The FLUKA code-based simulation results were consistent with the radiation detection.Conclusions The shielding calculation model based on FLUKA code can simulate the radiation field from proton and heavy ions.Among the secondary particles,secondary neutrons are the dominant component and the main concern of accelerator shielding design.In shielding calculation,the emphasis should be put on both beam intensity and energy.
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Objective To investigate the current situation of radiation protection in nuclear medicine diagnosis workplace.Methods The study was performed in 3 hospitals in northeast,north and central of China from February to December in 2013.The γ dose rate instrument was used to detect the workplace ambient dose equivalent rate of medicine preparation,leaching,packing,injection and imaging.Individual effective dose and equivalent dose were evaluated by photoluminescent dosimeter.Results The ambient dose equivalent rate was up to 1.92 mSv/h at repacking place and 1.2 mSv/h at injection place.The ambient dose equivalent rate of patients after injection was 5.36-240 μ,Sv/h.The hand equivalent dose was 0.01-0.02 mGy.Moreover,there were problems of staff route intersection,as well as the patients after injection staying in the public area.Conclusions Radiation workers should pay more attention to individual protection,and improve the operation proficiency to shorten the operation time.Furthermore,in order to protect public from unnecessary irradiation,there should be some changes in staff route and patients administration.
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Objective To study the effect of radon exhalation on the external dose model for building material,so as to provide the scientific and precise assessment of external radiation exposure hazard.Methods The mechanism of exhalation of radon from building material was analyzed,mathematical model of correction factor for the effect of radon exhalation was derived and resolved by Matlab program and the relationship between correction factor and diffusion length,surface emanation coefficient and thickness of building material was discussed.The absorbed dose rate induced by several classical building materials was calculated and compared.Results The radon exhalation correction factor was independent of diffusion length and thickness of building material in most cases.Negative correlation was found between radon exhalation correction factor and radon surface emanation coefficient.Radon exhalation correction factor numerically equals to '1-radon surface emanation coefficient'.The relative percentage deviation between absorbed dose rate induced by several classical building materials was in the range of 2.23％-10.02％,for both corrected and uncorrected radon exhalation effects.Conclusions Radon exhalation from building material has a certain effect on external dose model for building material,which should attract attention.It is important to conduct the correction for external dose model by introducing ‘1 -radon surface emanation coefficient’ as the radon exhalation correction factor,in order for the scientific assessment and control of external radiation exposure hazards from building materials.