23 results about "Activity concentration" patented technology

The invention relates to the technical field of accelerator radiation protection, in particular to an induced radioactivity assessment method suitable for a mobile accelerator component. Monte Carlo simulation software is adopted to calculate radionuclide activity concentrations in different areas of an acceleration component before movement, then radionuclides serve as radiation source items, and three-dimensional space dose rate distribution of the component and the periphery when the position of the accelerator component changes is obtained through calculation. And rapid evaluation of the received dose of the staff during maintenance and rapid formulation of a maintenance scheme are realized. The method effectively promotes the application of the Monte Carlo method in the field of radiation protection of the accelerator, provides reference for radiation hazards to workers, and provides technical support for radiation protection and operation maintenance of the accelerator.

The invention provides a radiation dose evaluation method and system based on radiation environment monitoring data, and relates to the technical field of radiation environment monitoring. The method comprises the following steps: recording the release source intensity of each nuclide at a release point, drawing an equally-spaced concentric arc grid, calculating the release height by combining the ground and the release point air pressure, air density and gravitational acceleration, and assigning nuclide hazard factors to the outer edge of the grid; according to the distance between the release point and the outer edge of the grid, diffusion parameters are calculated by using an empirical formula, a Gaussian atmospheric diffusion model is constructed in combination with the release height and hazard factors, and nuclide air activity concentration is output; comparing the total effective dose with a preset threshold value to calibrate a diffusion range, quantifying the radiation dose per unit area, and dividing atmospheric stability coefficient grades; constructing three types of correction factors of release height, material state and weather condition, establishing a comprehensive dose evaluation formula by combining the unit area radiation dose and the correction factors, comparing a preset threshold value, judging a risk level, and performing periodic updating.

The present disclosure relates to a method for monitoring failure of coated particles in fuel elements in a core of a pebble-bed high-temperature gas-cooled reactor, which is related to the technical field of nuclear reactor engineering and includes the following steps: S11, calculating an inventory of a short-lived noble gas fission nuclide; S12, obtaining a ratio of a release rate to a birth rate of the short-lived noble gas fission nuclide based on a temperature of the fuel elements using a Booth diffusion and release model; S13, deriving a theoretical expression for an activity concentration of the short-lived noble gas fission nuclide in a primary circuit using a migration model of the nuclide in the primary circuit; S14, obtaining an experimental measurement value of the activity concentration of the short-lived noble gas fission nuclide in the primary circuit at a sampling moment by gas sampling; S15, optimally calculating a failure fraction of the coated particles in the fuel elements and a share of uranium contamination in the matrix graphite in the core based on the theoretical expression and the experimental measurement value. The present disclosure can provide key parameters for the performance and status of the fuel elements in the core, which are required for radiation safety studies, source term calculations and accident analysis of the pebble-bed high-temperature gas-cooled reactor.

The application discloses a kind of nuclear power plant power operation mode pipeline deposition source item background evaluation system and method, this method includes: S1.the activity ratio ε of activated corrosion product nuclide i is measured i , pipe surface dose rate D and the activity concentration C of nuclide i in primary coolant i ;S2.modeling and calculating the effective dose rate D of pipe surface by activated corrosion product in deposition contribution d ;S3.modeling and calculating the surface activity of activated corrosion product nuclide i in deposition source item;S4.modeling and calculating the total count rate CPS or total absorbed dose rate Gy / h of pipeline deposition source item to detector.The nuclear power plant power operation mode pipeline deposition source item background evaluation system and method of the application, the activity ratio of activated corrosion product nuclide i, pipe surface dose rate, the activity concentration of nuclide i in primary coolant are measured, modeling and calculating the total count rate CPS or total absorbed dose rate Gy / h of pipeline deposition source item to detector, realize the evaluation of deposition source item background under the power operation mode of unit.

The present application relates to the technical field of accelerator radiation protection, and particularly relates to a method for evaluating induced radioactivity suitable for mobile accelerator components; the method is characterized in that the activity concentration of radionuclides in different regions of the accelerator components before movement is calculated by using Monte Carlo simulation software, then the radionuclides are taken as radiation source terms, and the three-dimensional spatial dose rate distribution of the components and the periphery when the position of the accelerator components changes is calculated, so that the rapid evaluation of the dose received by the staff during maintenance and the rapid formulation of the maintenance scheme are realized. The method effectively promotes the application of the Monte Carlo method in the field of accelerator radiation protection, provides a reference for the radiation hazards received by the staff, and provides technical support for the radiation protection and operation and maintenance of the accelerator.

This invention discloses a 89 SrCl2 solution 90 Mass spectrometry methods for measuring Sr activity concentration include: obtaining the concentration calculated during mass spectrometry measurement. 90 The formula required for the molar concentration of Sr is to use D in equation (12). F90 D E D was obtained by mass spectrometry; the analyte was calculated using equation (12). 89 SrCl2 solution 90 The molar concentration of Sr is then calculated. 90 The mass concentration of Sr; according to 90 Calculation of Sr mass concentration 90 The activity concentration of Sr; this invention eliminates the need for strontium and yttrium separation, simplifying the operation process, reducing personnel radiation exposure, and allowing direct calculation based on measurement data. 90 Sr activity concentration eliminates the intermediate strontium and yttrium activity concentration conversion process, avoiding 90 The impact of Y recovery rate on the final measurement results.

The invention discloses a method for detecting the activity concentration of 238U and 232Th in trona ore, and belongs to the technical field of inspection and detection. Comprising the following steps: S1, sample processing; grinding the trona ore at the temperature of less than 50 DEG C until the particle size is less than 100 meshes, drying, and immediately sealing and storing; s2, sample digestion; the method comprises the following steps: wetting a sample with water, dropwise adding hydrochloric acid until a violent reaction is stopped, evaporating to dryness, sequentially adding concentrated hydrochloric acid, concentrated nitric acid, hydrofluoric acid and perchloric acid, carrying out stepped heating dissolution and acid removal, and finally extracting with nitric acid to a constant volume; s3, measuring the mass concentration of the solution by a mass spectrometry method; measuring the mass concentration of 238U and 232Th in the solution by adopting inductively coupled plasma mass spectrometry and taking Re as an on-line internal standard; s4, calculating a result; and obtaining the activity concentration of the natural radionuclides 238U and 232Th in the measured trona ore through the mass-activity relation coefficient of the radionuclides.

The application discloses a kind of radioactive process system leakage monitoring methods, according to the dose rate level value of acquisition judges whether leakage occurs;And according to the typical nuclide spectrum analysis data of acquisition inversion leakage source item in the space to be measured, the activity concentration of typical radioactive nuclide in the space to be measured is obtained.The radioactive process system leakage monitoring method of the application strictly monitors leakage situation, so that relevant personnel can respond in time.The application also discloses a kind of radioactive process system leakage monitoring equipment.

The embodiment of the invention provides a method for determining the sorting limit value of a radioactive substance, and the method comprises the following steps: determining a plurality of activity concentration ratios among nuclides in the radioactive substance, and obtaining the calculation boundary activity concentration of each nuclide according to each activity concentration ratio. And preparing a plurality of samples corresponding to each calculated boundary activity concentration. And obtaining the corresponding radioactivity measurement values of each sample at different positions. And obtaining a sorting limit value according to each radioactivity measurement value and each calculated boundary activity concentration. The determination method provided by the embodiment of the invention is relatively low in cost and relatively high in analysis efficiency.

This invention relates to a method for calculating the minimum detectable activity concentration of a radionuclide, comprising the steps of: obtaining the source peak detection efficiency of a selected radionuclide; obtaining the background count of the selected radionuclide; calculating the minimum detectable activity concentration of the selected radionuclide; and forming a fitting formula based on the minimum detectable activity concentration of the selected radionuclide to calculate the minimum detectable activity concentration of other radionuclides in the same nuclear energy system. This invention also provides a system for calculating the minimum detectable activity concentration of a radionuclide. Using the method and system described in this invention, the minimum detectable activity concentration of other radionuclides in the nuclear energy system can be quickly calculated using the fitting formula to meet the accuracy requirements for engineering applications, reducing the workload of simulation calculations and saving manpower and time costs.

The present application discloses a method and device for determining the alarm threshold of the dose rate in a fresh air room, and a method and system for determining the alarm threshold of a monitoring instrument, which relate to the technical field of the nuclear industry. The method for determining the alarm threshold of the dose rate in a fresh air room comprises: obtaining the average activity concentration of the radioactive nuclide i in the main control room during the initial release period T3 of the first accident, and determining the first external exposure dose rate limit D allowed in the main control room based on the average activity concentration. ex,l ; According to the first external radiation dose rate limit D ex,l , determine the first alarm threshold D of the fresh air indoor dose rate vent,l According to the embodiment of the present application, the method can reasonably determine the first alarm threshold value D of the dose rate in the fresh air room. vent,l , is the first alarm threshold D of the subsequent fresh air indoor dose rate vent,l , which provides the possibility to reasonably determine the alarm threshold of the monitoring instrument.

The invention discloses an iodine filter box reuse performance evaluation method. The method comprises the following steps: obtaining an environment temperature W, an environment humidity S and an activity concentration A of radioactive iodine of a workplace; converting the activity concentration A of radioactive iodine into the mass concentration rho of non-radioactive methyl iodide according to a formula; replacing radioactive iodine with non-radioactive methyl iodide gas, making a simulation environment according to the environment temperature W, the environment humidity S and the mass concentration rho, and testing the unused iodine filter box in the simulation environment to obtain the protection time Tmax of the iodine filter box; obtaining a first preset value according to the protection time Tmax; obtaining the accumulated use duration T1 of the iodine filter box to be evaluated in the workplace; the T1 is compared with a first preset value; when T1 is smaller than a first preset value, it is judged that the iodine filter box is reusable. According to the iodine filter box reuse performance evaluation method, whether the iodine filter box can be reused or not can be scientifically and systematically evaluated, and resource saving and cyclic utilization of the iodine filter box can be promoted.

The invention provides an intrinsic measurement method and device for thorium. The intrinsic measurement method comprises the following steps: recording detection time of alpha particles released by decay of to-be-measured < 220 > Rn gas and daughter < 216 > Po in a measurement chamber; performing Bayesian analysis according to the detection time, and calculating a 220Rn-216Po true coincidence count; and according to the < 220 > Rn count, the < 216 > Po count and the < 20 > Rn-< 216 > Po true coincidence count, calculating the activity concentration of the thorium. According to the intrinsic measurement method and device, two alpha particles emitted by successive decay of 220Rn and 216Po are utilized to form an intrinsic characteristic conforming to a signal, so that the detection efficiency is reduced; the detection efficiency is reduced, so that the method and the device do not need to measure the detection efficiency of the standard scale from the outside, and have self-traceability; meanwhile, the influence of the change of the detection efficiency on the measurement accuracy is avoided, and the measurement accuracy is improved.

The application discloses a radioactive waste classification detection method, comprising the following steps: S10: obtaining weight information and density information of radioactive waste; S20: obtaining a nuclear signal of total activity of the radioactive waste to be detected, integrating the weight information, the density information and the nuclear signal of the total activity to obtain total activity concentration of the radioactive waste; S30: judging whether the total activity concentration exceeds an upper limit value range requirement of a very low radioactive level, and performing classification; S40: obtaining a nuclear signal of at least one nuclide activity of the radioactive waste to be detected, integrating the weight information, the density information and the nuclear signal of the nuclide activity to obtain nuclide activity concentration of at least one nuclide of the radioactive waste; and S50: judging whether the nuclide activity concentration exceeds the upper limit value range requirement of the very low radioactive level, and performing classification. The application can quickly and in batches classify radioactive waste through total amount detection and judgment for primary classification and nuclide detection and judgment for secondary classification, and is convenient for subsequent waste treatment.

The embodiment of the present application provides a detection method for the activity concentration of tritium gas, which comprises the following steps: controlling the first sub-channel to be in communication with the first gas inlet channel, controlling the to-be-detected gas to enter the first sub-channel from the first gas inlet channel, detecting the activity concentration of tritium gas in the to-be-detected gas by using the first detection device, judging whether the activity concentration of tritium gas in the to-be-detected gas is less than a first threshold value, wherein the first threshold value is the minimum value of the first activity concentration, and determining the on-off state of the second sub-channel and the first gas inlet channel according to the judgment result. The detection method of the embodiment of the present application can select the corresponding sub-channel and the corresponding detector for detection according to the size of the activity concentration of tritium gas, so that the tritium gas with the activity concentration greater than the first threshold value can be prevented from entering the second sub-channel and the second detection device, that is, the tritium gas with the activity concentration greater than the first threshold value can be prevented from causing pollution to the second detection device and the second sub-channel, and thus the detection accuracy of the tritium monitor is improved.

The present application relates to the technical field of nuclear radiation detection, and particularly relates to a radon-tritium separation discrimination measurement device and method in a high-radon environment, which comprises a first measurement channel, a second measurement channel, a third measurement channel and a fourth measurement channel, and respectively measures a first current value I1, a second current value I2, a third current value I3 and a fourth current value I4, and respectively calculates the activity concentration C Rn of radon Rn, the activity concentration C HTO of tritiated water HTO and the activity concentration C HT of tritium gas HT according to the first current value I1, the second current value I2, the third current value I3 and the fourth current value I4. The present application focuses on the characteristics of indoor environment, considers the influence of indoor radon on tritium measurement, adopts a series-parallel differential mode, avoids the low detection limit influence caused by multi-channel measurement, is more suitable for low-concentration-level environment measurement, saves the need of purging gas, and simplifies the device structure.

The invention provides a PING monitor prototype radon chamber test method, device and equipment and a storage medium, and aims to solve the technical problem of poor performance evaluation accuracy of a PING monitor prototype. The test method comprises the following steps: extracting gas to be tested; S10, acquiring the activity concentration of radon gas and the equilibrium equivalent concentration of radon daughters in the gas to be tested; a gas detection step S20, detecting to obtain an alpha particle net counting rate and a first beta particle net counting rate; s30, determining a radon daughter response coefficient, and determining a radon daughter response coefficient and compensation coefficient curve; s40, the response coefficient of the radon daughters is compensated, and the response coefficient of the radon daughters after compensation is obtained; a second beta particle net counting rate detection step S50, obtaining a second beta particle net counting rate; a radon gas response coefficient detection step S60, determining a radon gas response coefficient; and a monitor prototype evaluation step S70, evaluating the performance of the monitor prototype according to the compensated radon daughter response coefficient and the radon gas response coefficient.

The invention relates to the technical field of accelerator radiation protection, in particular to an induced radioactivity assessment method suitable for a particle accelerator beam tube. The method aims at providing reference for radiation protection and operation mode optimization of an accelerator and maintenance plan making for workers by evaluating induced radioactivity of an accelerator beam tube in a multi-cycle operation mode. According to the method, the operation mode of an accelerator is considered, Monte Carlo simulation software is adopted, and the radionuclide activity concentration, the focus dose rate and the dose received by a worker during long-time maintenance of a focus in a beam tube in different operation periods and different cooling time are calculated through an activation formula in a multi-period operation mode; and future induced radioactivity is predicted, so that application of the Monte Carlo method in the field of radiation protection of the accelerator is effectively promoted, and technical support is provided for radiation protection and operation maintenance of the accelerator.

The present application relates to a method and system for measuring radioactive waste barrels outside a nuclear power plant, the method comprising: obtaining measurement parameters of a waste barrel to be measured; determining the detection efficiency of equivalent nuclides and calculation data according to the measurement parameters; calculating the activity concentration of all easy-to-measure gamma nuclides and all difficult-to-measure gamma nuclides according to the detection efficiency of equivalent nuclides and the calculation data; and evaluating whether the waste barrel to be measured meets the clean control requirements according to the activity concentration of all easy-to-measure gamma nuclides and all difficult-to-measure gamma nuclides. The present application can significantly improve the accuracy, reliability and automation level of barrel-out measurement, and provides a powerful technical means for the clean control of radioactive waste.

The present invention discloses a method for measuring the deposition source term at a pressurized water reactor nuclear power plant, comprising the following steps: initially establishing a deposition point arrangement plan; determining the radioactive nuclides to be measured, obtaining the activity concentration of the radioactive nuclides deposited on the inner wall of the pipe at each measurement point, and sorting the measurement points according to the activity concentration; determining the installation conditions of the proposed deposition source term online measurement device, measuring the on-site space dimensions of each measurement point, and excluding measurement points that do not meet the installation conditions; determining whether there are additional measurement points and priority factors; if there are no additional measurement points and priority factors, the order of the deposition point arrangement plan remains unchanged; if there are additional measurement points and / or priority factors, the additional measurement points are added and the measurement points are reordered; and according to the number of deposition source term online measurement devices, redundant measurement points are excluded, and finally the deposition point arrangement plan is determined. The deposition point arrangement method of the present invention is highly operable and can provide a reference for determining the on-site gamma spectrum online measurement point arrangement plan for various nuclear power plants.

The application provides a capability verification liquid sample stability test method and related equipment, and relates to the technical field of radionuclide analysis.The method comprises the following steps: randomly drawing a preset number of to-be-processed liquid samples from a liquid sample, determining a stability test parameter, placing the to-be-processed liquid samples into corresponding test environments for processing, obtaining to-be-tested liquid samples, and the stability test parameter at least comprising a temperature.An analysis is performed according to a standard analysis method to obtain a measurement average value of a radionuclide, the measurement average value is corrected to a preset test date according to a half-life of the radionuclide to obtain a corrected average value, and a stability conclusion of the liquid sample under the stability test parameter is obtained according to the corrected average value.The method can test factors affecting the stability of the capability verification liquid sample, avoid the influence of these factors during the analysis of the activity concentration of the radionuclide, and improve the reliability of the analysis result of the activity concentration of the radionuclide.

The invention discloses a method for measuring the content of 227Ac in a 225Ac solution, and belongs to the technical field of nuclide analysis. Comprising the following steps: providing a 227Ac diluent solution with known activity concentration; equally dividing a to-be-detected 225Ac solution into two equal parts, adding a quantitative 227Ac diluent solution into one part as a labeled sample, and taking the other part as a non-labeled sample; respectively measuring the isotope ratios of 227Ac / 225Ac in the two samples; based on the principle of isotope dilution mass spectrometry, by comparing two ratios and combining the sampling amount, the diluent solution adding amount and the activity concentration of the diluent solution, a material balance relation is constructed, and the activity concentration of 227Ac and 225Ac in the to-be-detected solution is obtained. According to the invention, absolute measurement is realized through the principles of high-sensitivity measurement, micro-sampling and an isotope dilution method of a mass spectrometer; and by combining a material balance equation, the method for rapidly and accurately measuring the content of 227Ac in the 225Ac solution in a trace amount can be realized.

The embodiment of the invention provides a method for determining a sorting limit value of a radioactive substance. The method comprises the following steps: acquiring activity concentration of each nuclide contained in the radioactive substance and a classification activity concentration guide value corresponding to each nuclide; and obtaining the sorting limit total radioactivity specific activity according to each activity concentration and each classification activity concentration guide value. According to the determination method provided by the embodiment of the invention, the classification sorting efficiency of the radioactive substances can be improved.