Objective To preliminarily study and establish a method for measurement of the transuranic nuclide ²⁴¹Am in fecal samples, and to provide technical support for internal radiation monitoring of staff. Methods Fecal samples were collected with a self-made stool sampler and treated with a self-made carbonization and ashing furnace. DGA resin was used to separate and purify ²⁴¹Am from fecal samples. With ²⁴³Am as the tracer, the orthogonal method was used for condition optimization. Results The optimum conditions for separation and purification were: the acidity of HNO₃ added into the column, 6 mol/L; column flow rate, 0.6 mL/min; and the volume of analytical solution,12 mL. The method based on inductively coupled plasma mass spectrometry showed a detection limit of 9.79×10⁻⁴ Bq for ²⁴¹Am in fecal samples, which was satisfactory and feasible. Conclusion This method fills the vacancy of ²⁴¹Am measurement in fecal samples to some extent, which is of practical significance for internal radiation monitoring and protection for analysts.

Objective To establish a method for measurement of ²³⁹Pu in fecal samples based on inductively coupled plasma-mass spectrometry (ICP-MS), and to provide a novel method for assessing the internal exposure of workers. Methods Fecal samples were collected from workers and labeled. The samples were pretreated with carbonization ashing and microwave digestion devices, purified on TEVA resin, and measured using ICP-MS. Results The detection limit of ²³⁹Pu in fecal samples based on ICP-MS was 1.91 × 10⁻⁴ Bq. Conclusion In the routine monitoring of class S substances characterized by a 5 μm aerodynamic diameter during 12 months, the committed effective dose corresponding to the detection limit is 0.17 mSv. This value meets the requirements of relevant national standards and ICP-MS can be used as a novel means for accurate evaluation of internal exposure for workers.

Objective To develop a simplified phantom for the calibration of whole-body counters. Methods A simplified phantom design method for the calibration of whole-body counters was established based on the process and method of calibrating whole-body counters. By using the established method and Monte Carlo method, a simplified phantom including the total body, thyroid, lungs, and gastrointestinal tract was designed to calibrate the ORTEC-Stand FAST II whole-body counter. The simplified phantom was compared with the BOMAB phantom through experimental measurements. Results Within the range of 50 keV to 2 MeV, for rays of the same energy in the same organ of the simplified phantom and BOMAB phantom, the simulated data of detection efficiency by whole-body counting showed an error within 5%, and the experimental measurements showed an error within 10%. Conclusion We developed a simplified phantom for the calibration of the whole-body counter, demonstrating the feasibility of using the simplified phantom instead of a physical body phantom for whole-body counter calibration, which can greatly facilitate whole-body counter calibration for internal radiation monitoring.

Objective To evaluate the migration of plutonium aerosol caused by α recoil. Methods In this paper, the recoil deposition and Brownian motion of plutonium-containing nanoaerosols were simulated by Monte Carlo method. The recoil angle and the vertical first landing time of Brownian motion in the process of settling were sampled, and then the lateral displacements of Brownian motion were sampled to determine the final settling position of aerosol. Results For aerosols with particle sizes of 10-50 nm, the maximum migration distance of a single recoil settling was 1.39 μm. Brownian motion increased the migration capacity. Although there was a high likelihood that aerosols settled within 100 μm, there remained a slight probability of long-term suspension in the air. Conclusion The α recoil is one of the mechanisms of plutonium aerosol migration. An important mechanism for long-distance migration of nanoaerosols is that Brownian motion after recoil may cause them to suspend for a long time.

Objective To calculate the absorbed dose of ⁹⁰Y TheraSphere in the pancreas and the surrounding sensitive organs after the administration in the treatment of pancreatic cancer through the establishment of an individual voxel model, and to provide technical support for the clinical application of ⁹⁰Y TheraSphere in the treatment of pancreatic cancer. Methods An individualized voxel model was constructed in Geant4 software based on the CT images of the patient. 12 monoenergetic electron specific absorption fractions (SAFs) in the range of 0.01 to 1 MeV were calculated and validated against the ICRP data. The model and method were used to calculate the absorbed doses in the target organs under uniform and nonuniform distribution of ⁹⁰Y microspheres in the pancreas. Results The relative errors between the SAF values calculated based on the individualized voxel model and the ICRP data after mass calibration were less than 3.89%. When ⁹⁰Y was uniformly distributed in the pancreas, the absorbed dose in the pancreas was 4.69 × 10⁻⁷ Gy/Bq; the absorbed doses in the liver, kidneys, and spleen were 6.15 × 10⁻¹², 6 × 10⁻¹², and 1.65 × 10⁻¹¹ Gy/Bq, respectively. When ⁹⁰Y was distributed within the tumor, the absorbed dose in the tumor was 6.69 × 10⁻⁶ Gy/Bq and the absorbed dose in normal pancreas was 5.72 × 10⁻⁸ Gy/Bq. The fitted relationship between tumor volume V and administered activity A at the prescribed dose of 120 Gy was quadratic, with relatively low activity required for concentrated administration in the center of the tumor. Conclusion The Monte Carlo dose calculation method based on individual voxel model accurately predicted the absorbed doses in the surrounding sensitive organs (liver, kidneys, and spleen) when ⁹⁰Y TheraSphere was used to treat pancreatic cancer. These results and the analysis of the factors affecting the drug delivery activity will provide data support for the clinical research of ⁹⁰Y TheraSphere in pancreatic cancer.

Objective To address the radioactive contamination of wounds caused by transuranic nuclides, wound radiation imaging based on coded aperture imaging technology was investigated. Methods By simulating multiple source terms using Monte Carlo method, the differences in imaging performance between two image reconstruction algorithms under near-field conditions were compared. The effects of detector pixels and detection plane pixels on image resolution were investigated. Results The imaging system was simulated based on the designed dimensions. The simulated imaging field of view was 89.4 mm × 89.4 mm and the simulated angular resolution was 1.98°. Based on the comparison of the average width at half height of the reconstructed point sources under different conditions, it was found that increasing the number of pixels in the detector and detection plane optimized the angular resolution but significantly prolonged the Monte Carlo simulation time. Conclusion According to the simulation results, the parameters of the imaging system can be used to effectively image radioactive contamination. Our results provide methodological support for the measurement of wound contamination caused by transuranic nuclides, and lay the foundation for the development of wound contamination imaging detection systems in the future.

Objective To investigate the separation efficiency of three physical separation methods for gaseous ¹⁴C, namely membrane separation, adsorption separation, and low-temperature separation, to screen for the optimal separation method, and to provide a reference for the separation and enrichment of ¹⁴CO₂ in online monitoring of ¹⁴C. Methods The experimental plan was designed, and three devices were constructed for separation and purification experiments. The purity, recovery rate, and separation time of CO₂ separated by the three methods were analyzed. Results All the three methods achieved the separation of CO₂. Under certain conditions, 20 mL of sample gas was obtained. The separation time of membrane separation method was 0.5 hour, CO₂ gas with a sample purity of 0.0486%-0.0488%VOL was obtained, and the recovery rate was only 2.5%−12.7%. The separation time of MOF material adsorption separation method was 24−30 hours, CO₂ gas with a sample purity of 20.2%−47.5%VOL was obtained, and the recovery rate was 57.3%−63.2%. The separation time of low-temperature separation method was 3 hours, CO₂ gas with a sample purity of 96.0%−99.8%VOL was obtained, and the recovery rate was 81.5%−92.5%. Conclusion The purity and recovery rate of CO₂ with low-temperature separation method were higher compared to membrane separation method and MOF material adsorption separation method. The separation time of low-temperature separation method was moderate. The low-temperature separation method can provide a reference for the enrichment technology of ¹⁴C in the rapid measurement of gaseous ¹⁴C effluent.

Objective Nowadays, radioactive xenon isotopes, including ^131mXe, ^133mXe, ¹³³Xe, and ¹³⁵Xe, are primarily released into the atmosphere through various reactor operation and major accidents of reactors. To improve the online monitoring capability of xenon in nuclear facilities and their gaseous effluents, a highly sensitive online xenon monitoring system was developed to monitor, warn, and alarm the activity concentration of radioactive xenon. Methods The online monitoring system for radioactive xenon gas in nuclear facilities was established using xenon membrane separation and concentration, xenon high-efficiency selective adsorption, and low-background gamma-ray spectrometry analysis methods. Results Under the operation mode of one-hour sampling and one-hour measuring, the minimum detectable activity concentration of the radioactive xenon online monitoring system for ¹³³Xe was approximately (1.43 ± 0.03) Bq/m³. Conclusion This system can be effectively used for online monitoring of xenon activity concentration in nuclear facilities such as nuclear power plants and isotope production reactors, as well as in gaseous effluents. It helps improve the safety level of personnel, the environment, and nuclear facilities.

Objective To design a mobile personnel radiation protection equipment for operation in environments with high radiation such as spent fuel reprocessing plants, to achieve simultaneous protection against γ radiation, neutron radiation, and radioactive aerosol, to reduce the internal and external exposure dose of radioactive workers, and to meet the requirement of operation for two hours. Methods The core parts of the mobile personnel radiation protection equipment included a shielding chamber and a respiratory maintenance system. An automated chassis was used for the movement and lifting of the shielding chamber. MCNP software was used to simulate and calculate the protective effects of shielding chamber made of different materials and material thicknesses. Experimental verification of the shielding chamber design was conducted. Mathematical models were established to describe the variations in the content of various gases in the chamber with personnel operation time. A respiratory maintenance system, a harmful gas absorption device, and an automated mobile chassis were designed. Results The shielding chamber made of polyethylene with a thickness of 80 mm achieved an 80% neutron shielding rate. The respiratory maintenance system could support workers for 2 hours of operation inside the equipment. The mobile chassis allowed operation of the equipment with one person. Conclusion This mobile personnel radiation protection equipment can solve the problem in simultaneous protection against γ radiation, neutron radiation, and radioactive aerosol. The equipment can provide radiation protection for radioactive workers, reduce exposure dose, and reduce personnel burden. This system provides technical means for the operation and maintenance of equipment in high-radiation sites such as spent fuel reprocessing plants.