Measuring device and measuring method for nuclide leakage path factors under high-level liquid waste accident condition

Abstract

The invention discloses a measurement device and a measurement method for nuclide leakage path factors under a high-level liquid waste accident condition. The measurement device comprises a release module, a gas phase adjustment and mixing module, a migration path simulation module and a nuclide collection and measurement module, the release module comprises a sample container, a program temperature control heating furnace and a carrier gas supply unit, and the carrier gas supply unit comprises a first mass flow controller for adjusting the carrier gas flow; the gas phase adjusting and mixing module comprises a steam generator and a preheating mixing chamber, and the preheating mixing chamber is used for mixing released product gas flow from the releasing module with steam flow output by the steam generator; the migration path simulation module comprises a pipeline and / or a box body and is used for simulating a typical geometric migration path of nuclide and bearing the flowing and deposition process of mixed airflow; the nuclide collection and measurement module is used for sequentially intercepting aerosol-state nuclides, condensing and dissolving condensable gaseous nuclides and chemically absorbing residual gaseous nuclides.

Description

A device and method for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions. Technical Field

[0001] This invention relates to the field of nuclear facility safety analysis, and specifically to a measuring device and method for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions. Background Technology

[0002] The five-factor method is a commonly used method for calculating source terms in non-reactor nuclear facilities. This method uses five parameters multiplied together: MAR×DR×ARF×RF×LPF, where: MAR is the amount of waste liquid accumulated in the tank; DR is the damage ratio, the proportion released from the tank; ARF is the airborne release share; RF is the breathing share; and LPF is the leakage path factor, which takes into account the release path and is mainly affected by filtration and sedimentation.

[0003] Current methods mainly involve mixing a stable gaseous nuclide source with a controlled simulated atmosphere and studying deposition behavior through a constant-temperature pipeline. Another method involves directly heating the simulated waste liquid and introducing steam into a model space (such as a stainless steel tank), calculating the LPF by measuring the amount of nuclides inside and outside the tank.

[0004] Current technologies cannot accurately control and separate the environmental conditions of the migration path while simulating the actual waste liquid release process, thus limiting the accuracy and universality of LPF measurements. On the one hand, there is a lack of ability to accurately simulate and control the complex gas-phase chemical environment (such as high temperature, high humidity, and high concentration of nitric acid vapor) in high-level radioactive waste liquid accidents. On the other hand, there is a lack of in-depth understanding of the chemical speciation, deposition, and migration mechanisms of nuclides (such as ruthenium) under different gas-phase conditions (such as dry air, water vapor, and nitric acid-containing water vapor). Summary of the Invention

[0005] To achieve the above and other related objectives, this invention discloses a device for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions, comprising: a release module, a gas phase conditioning and mixing module, a migration path simulation module, and a radionuclide collection and measurement module connected sequentially along the gas flow direction; wherein, the release module includes: a sample container for containing simulated high-level radioactive waste, a programmable temperature controlled furnace for covering and heating the sample container to simulate the heating process under accident conditions, and a carrier gas supply unit for conveying the release products from the sample container, the carrier gas supply unit including a first mass flow controller connected to a carrier gas source for adjusting the carrier gas flow rate, the outlet of which is connected to the upper part of the sample container so that the carrier gas carries the release products from top to bottom; the gas phase conditioning and mixing module includes: a steam generator for generating water vapor and / or nitric acid-water vapor mixture, and a device connected to the outlet of the steam generator and connected to the gas phase control module. The release module outlet is connected to a preheating mixing chamber, which is used to mix and homogenize the release product gas flow from the release module with the steam flow output from the steam generator at a set temperature, thereby forming a gaseous environment with a predetermined chemical composition and physical state; the migration path simulation module includes a pipe and / or a box equipped with a temperature control unit, used to simulate the typical geometric migration path of nuclides from the nuclear facility waste liquid release point to the plant or environment under controllable temperature conditions, and used to carry the flow and deposition process of the mixed gas flow; the nuclide collection and measurement module includes, in sequence along the gas flow direction, a high-efficiency particulate filter, a condenser, and at least one gas scrubbing bottle containing absorbent liquid, used to sequentially intercept aerosol nuclides, condense and dissolve condensable gaseous nuclides, and chemically absorb residual gaseous nuclides, and to measure the total amount of nuclides passing through the migration path simulation module by quantitative analysis of the amount of nuclides recovered in each module.

[0006] Preferably, the migration path simulation module is composed of a linear reaction pipeline assembly and a box assembly connected in series. The outlet of the linear reaction pipeline assembly is sealed to the inlet of the box assembly. The linear reaction pipeline assembly is used to simulate the migration behavior of waste liquid along linear paths such as ventilation ducts and process pipelines after release. The box assembly is used to simulate mixing, condensation and wall effects in the process room and equipment compartment.

[0007] Preferably, the linear reaction pipeline assembly is composed of multiple sections of glass or stainless steel pipes with the same inner diameter, which are detachably connected by flanges. Each pipe section can be disassembled individually to facilitate the segmented recovery of deposited nuclides. The linear reaction pipeline assembly and the box assembly are housed in a constant temperature chamber. The constant temperature chamber has the function of adjusting and maintaining a uniform temperature within a predetermined temperature range to form a stable and uniform temperature field throughout the migration path.

[0008] Preferably, the steam generator includes two injection pumps for conveying deionized water and a nitric acid solution of a predetermined concentration, an atomizer for atomizing the liquid, and a steaming pipe connected to the atomizer and heated by a heating unit. The deionized water and nitric acid solution conveyed by the injection pumps are atomized into droplets in the atomizer and then enter the steaming pipe. In the steaming pipe, they are heated and vaporized to form water vapor or a nitric acid-water vapor mixture, and then output to the preheating mixing chamber under the carry gas controlled by the second mass flow controller.

[0009] Preferably, at least one gas scrubbing bottle in the nuclide collection and measurement module contains an aqueous sodium hydroxide solution, preferably with a concentration of 0.05~0.5 mol / L, for absorbing acidic gaseous nuclides; and the nuclide collection and measurement module further includes a second gas scrubbing bottle sequentially connected to the sodium hydroxide aqueous solution scrubbing bottle, the second gas scrubbing bottle containing a hydrochloric acid-ethanol mixed solution with a volume ratio of 1:1, serving as a safeguard unit to reduce and absorb ruthenium compounds and / or other high-valence gaseous nuclides that penetrate the preceding scrubbing bottle.

[0010] Preferably, the sample container is a cylindrical container made of quartz glass or Hastelloy C276, which has the characteristics of high temperature resistance and strong acid corrosion resistance. The programmable temperature-controlled heating furnace is a tubular resistance furnace with a temperature control system that can set multiple heating programs and has a temperature control accuracy of not less than ±1℃. Furthermore, the release module is equipped with a thermocouple inserted below the sample liquid surface to monitor the temperature of the simulated high-level radioactive waste liquid in real time, thereby realizing the controllable simulation of the radionuclide release rate and release stage.

[0011] Secondly, this invention discloses a method for measuring the radionuclide leakage path factor under high-level radioactive waste liquid accident conditions. Using the aforementioned measuring device, the method includes the following steps: system preparation and calibration, including: preparing a simulated high-level radioactive waste liquid containing the target radionuclide; adding the simulated high-level radioactive waste liquid to the sample container; without connecting the migration path simulation module, directly connecting the outlet of the release module to the inlet of the radionuclide collection and measurement module via a short path; heating the simulated high-level radioactive waste liquid in the sample container according to a predetermined temperature-controlled heating program, causing the target radionuclide to be released and collected by the radionuclide. With the entire measurement module recovered, quantitative analysis of the solutions and filters in each unit of the nuclide collection and measurement module was conducted to obtain the total release amount of the target nuclide and its release rate over time under the heating program. The leakage path factor measurement experiment, including the connection of the complete release module, gas phase conditioning and mixing module, migration path simulation module, and nuclide collection and measurement module, involved setting parameters such as the waste liquid heating program, gas phase composition ratio, carrier gas and vapor flow rate, and migration path temperature according to the target accident conditions. The constant temperature chamber and gas phase conditioning and mixing module were preheated to reach and stabilize at the set temperature before startup. The waste liquid heating program, along with various mass flow controllers and injection pumps, mixes the release products containing the target nuclide with water vapor and / or nitric acid-water vapor mixtures generated by the steam generator in the preheating mixing chamber. The mixture then sequentially passes through the migration path simulation module and the nuclide collection and measurement module until the simulated high-level radioactive waste liquid is completely dry and reaches the endpoint temperature. Afterward, carrier gas and steam are introduced for a purging time, and the experiment is stopped. Various samples from the migration path simulation module and the nuclide collection and measurement module are collected. The experimental results are analyzed and calculated, including: the pipes and / or... The chamber is subjected to zoned washing and / or soaking to recover deposited nuclides, and the solutions and filters of each unit in the nuclide collection and measurement module are digested. Inductively coupled plasma mass spectrometry or other appropriate analytical methods are used to quantitatively analyze the target nuclides in all samples, and the total amount of deposited nuclides D recovered from the migration path simulation module and the total amount of nuclides P recovered from the nuclide collection and measurement module are calculated. Based on this, the total recovery amount T = P + D is calculated. When the total recovery rate is within a preset allowable range, the leakage path factor of the nuclide under the accident condition is obtained according to the formula LPF = P / T.

[0012] Preferably, system preparation and calibration include: calibrating the nuclide release rate by directly connecting the release module to the nuclide collection and measurement module during the heating process and bypassing the migration path simulation module, and using the total release amount and release rate curve obtained from the calibration as the benchmark for the nuclide source term in the subsequent leakage path factor measurement experiment.

[0013] Preferably, in the leakage path factor measurement experimental step, the inner surface of the pipe and / or box in the migration path simulation module is treated sequentially as follows: first, it is rinsed multiple times with an acidic solution to recover soluble nuclide deposits; then, the pipe and / or box is immersed in an oxidation-reduction eluent containing bisulfite and alkaline components and kept for a predetermined time under heating and / or ultrasonic vibration conditions to dissolve insoluble or sparingly soluble nuclide deposits; the acidic washing solution and eluent are collected separately for quantitative analysis, so that soluble and insoluble nuclide deposits in the migration path are recovered stepwise, thereby improving the accuracy of the mass balance of deposited nuclides in the leakage path factor calculation.

[0014] By employing the aforementioned technical solutions, a real radionuclide release source term was simulated through the release module, while precise control of the chemical environment along the migration path was achieved through the gas-phase conditioning and mixing module. The migration path simulation module, which connects linear pipelines and the enclosure in series, allows for the simultaneous study of linear deposition behavior of radionuclides within the pipelines and condensation, mixing, and wall effects within the large space, providing a powerful tool for revealing the dominant mechanism of LPF (Limited-Fluid Activation). A comprehensive, hierarchical radionuclide collection and measurement module was designed, combined with a rigorous chemical processing procedure, ensuring radionuclide mass balance and making the final calculated LPF value highly reliable. Key operating parameters of the device can be independently adjusted, facilitating parameter sensitivity analysis. It can be widely used to assess radionuclide release risks under different accident sequences and facility designs, providing solid data support for safety reviews and the development of accident management measures. Attached Figure Description

[0015] The above and other features, advantages, and aspects of the embodiments of this disclosure will become more apparent from the accompanying drawings and the following detailed description. The drawings are provided for a better understanding of the invention and are not intended to limit the scope of this disclosure. In the drawings, the same or similar reference numerals denote the same or similar elements, wherein: Figure 1 is a schematic diagram of the principle of an embodiment of the present invention; Figure 2 is a schematic diagram of the principle of the release module and the gas phase regulation and mixing module in an embodiment of the present invention; Figure 3 is a schematic diagram of the migration path simulation module in an embodiment of the present invention; Figure 4 is a schematic diagram of the nuclide collection and measurement module in an embodiment of the present invention.

[0016] Reference numerals: 1. High-purity dry air source; 2. Heating furnace; 3. Sample container; 4. Simulated high-level radioactive waste liquid; 5. Nebulizer; 6. Stainless steel pipe; 7. Heating tape; 8. Preheating mixing chamber; 9. Air heater; 10. Linear piping assembly; 11. Box assembly; 12. High-efficiency particulate filter; 13. Condenser; 14. Primary gas scrubbing bottle; 15. Secondary gas scrubbing bottle; 16. Collection bottle. Detailed Implementation

[0017] The technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort are within the scope of protection of the present invention.

[0018] Referring to Figure 1, an embodiment of the present invention provides a measuring device for the radionuclide leakage path factor under high-level radioactive waste liquid accident conditions, comprising: a release module, a gas phase conditioning and mixing module, a migration path simulation module, and a radionuclide collection and measurement module connected sequentially along the gas flow direction; wherein, the release module includes: a sample container 3 for holding simulated high-level radioactive waste liquid 4, a programmable temperature controlled heating furnace 2 for covering and heating the sample container 3 to simulate the heating process under accident conditions, and a carrier gas supply unit for conveying the release products from the sample container 3, the carrier gas supply unit including a first mass flow controller connected to the carrier gas source for adjusting the carrier gas flow rate, the outlet of which is connected to the upper part of the sample container 3 so that the carrier gas carries the release products from top to bottom; the gas phase conditioning and mixing module includes: a steam generator for generating water vapor and / or nitric acid-water vapor mixed vapor, and a steam generator connected to the outlet of the steam generator and the release module The preheating mixing chamber 8, connected to the block outlet, is used to mix and homogenize the release product gas flow from the release module with the steam flow output from the steam generator at a set temperature, thereby forming a gaseous environment with a predetermined chemical composition and physical state. The migration path simulation module includes a pipe and / or a box equipped with a temperature control unit, used to simulate the typical geometric migration path of nuclides from the nuclear facility waste liquid release point to the plant or environment under controllable temperature conditions, and to carry the flow and deposition process of the mixed gas flow. The nuclide collection and measurement module includes, in sequence along the gas flow direction, a high-efficiency particulate filter 12, a condenser 13, and at least one gas washing bottle containing absorbent liquid, used to sequentially intercept aerosol nuclides, condense and dissolve condensable gaseous nuclides, and chemically absorb residual gaseous nuclides, and to measure the total amount of nuclides passing through the migration path simulation module by quantitative analysis of the amount of nuclides recovered in each module.

[0019] Preferably, the migration path simulation module is composed of a linear reaction pipeline assembly and a box assembly 11 connected in series. The outlet of the linear reaction pipeline assembly is sealed to the inlet of the box assembly 11. The linear reaction pipeline assembly is used to simulate the migration behavior of waste liquid along linear paths such as ventilation ducts and process pipelines after release. The box assembly 11 is used to simulate mixing, condensation and wall effects in the process room and equipment compartment.

[0020] Preferably, the linear reaction pipeline assembly is composed of multiple glass tubes or stainless steel tubes 6 with the same inner diameter connected detachably by flanges. Each tube segment can be disassembled individually to facilitate the segmented recovery of deposited nuclides. The linear reaction pipeline assembly and the box assembly 11 are set up in a constant temperature chamber. The constant temperature chamber has the function of adjusting and maintaining a uniform temperature within a predetermined temperature range to form a stable and uniform temperature field throughout the migration path.

[0021] Preferably, the steam generator includes two injection pumps for conveying deionized water and a nitric acid solution of a predetermined concentration, an atomizer 5 for atomizing the liquid, and a steaming pipe connected to the atomizer 5 and heated by a heating unit. The deionized water and nitric acid solution conveyed by the injection pumps are atomized into droplets in the atomizer 5 and then enter the steaming pipe. In the steaming pipe, they are heated and vaporized to form water vapor or a nitric acid-water vapor mixture, and then output to the preheating mixing chamber 8 under the carry gas controlled by the second mass flow controller.

[0022] Preferably, at least one gas scrubbing bottle in the nuclide collection and measurement module contains an aqueous sodium hydroxide solution, preferably with a concentration of 0.05~0.5 mol / L, for absorbing acidic gaseous nuclides; and the nuclide collection and measurement module further includes a second gas scrubbing bottle sequentially connected to the sodium hydroxide aqueous solution scrubbing bottle, the second gas scrubbing bottle containing a hydrochloric acid-ethanol mixed solution with a volume ratio of 1:1, serving as a safeguard unit to reduce and absorb ruthenium compounds and / or other high-valence gaseous nuclides that penetrate the preceding scrubbing bottle.

[0023] Preferably, the sample container 3 is a cylindrical container made of quartz glass or Hastelloy C276, which has the characteristics of high temperature resistance and strong acid corrosion resistance. The programmable temperature-controlled heating furnace 2 is a tubular resistance furnace with a temperature control system that can set multiple heating programs and has a temperature control accuracy of not less than ±1℃. Furthermore, the release module is equipped with a thermocouple inserted below the sample liquid surface to monitor the temperature of the simulated high-level radioactive waste liquid 4 in real time, thereby realizing the controllable simulation of the radionuclide release rate and release stage.

[0024] In a preferred embodiment of the present invention, the measuring device provided by the present invention is a modular, end-to-end integrated system, the core of which consists of four modules with clearly defined functions that work in series and cooperate with each other.

[0025] Referring to Figure 2, the release module is used to physically simulate the accident process of the HLLW storage tank and provide a stable and controllable source of nuclide release. Specifically, it includes: Sample container 3: a cylindrical container made of quartz glass or Hastelloy C276 (inner diameter 30mm, height 200mm, volume approximately 140mL), possessing excellent high-temperature resistance and corrosion resistance. Programmable temperature controlled heating furnace 2: a tubular resistance furnace with a temperature control accuracy of ±1℃ and a maximum operating temperature of 500℃. The heating program can be set in segments to simulate the slow temperature rise (heating rate 1℃ / min) caused by decay heat or an accident, followed by a rapid drying stage. Carrier gas supply unit: including a high-purity dry air source 1, a pressure reducing valve V1, and a first mass flow controller MFC1. The MFC1 has a range of 0-0.5NL / min and an accuracy of ±1%FS. The carrier gas is introduced from the top of sample container 3, carrying the released gaseous nuclides and vapors. Temperature monitoring: Two K-type thermocouples are directly inserted below the liquid surface in sample container 3 to monitor the temperature of the waste liquid in real time, including thermocouple T1 and thermocouple T2.

[0026] Referring to Figure 2, the gas phase conditioning and mixing module is used to precisely generate and regulate the chemical composition and physical state of the gas phase environment entering the migration path. Specifically, it includes: a steam generator; a liquid delivery system: two independent precision syringe pumps, including syringe pump P1 and syringe pump P2, used to deliver deionized water and a predetermined concentration of nitric acid solution (e.g., 2 mol / L), with a flow rate range of 0-100 μL / min; an ultrasonic nebulizer 5: the liquid is atomized using an ultrasonic nebulizer 5, and the droplets then enter a stainless steel tube 6 wrapped with a heating cable 7 (maintained at 150-200°C), instantly vaporizing into steam. A thermocouple T3 is installed inside the stainless steel tube 6 for temperature measurement, and a thermocouple T4 is installed between the heating cable 7 and the stainless steel tube 6 for temperature measurement; and a carrier gas supply: a second mass flow controller (MFC2) provides another stream of dry air to carry the generated steam, with a flow rate range of 0-1.0 NL / min. Preheating Mixing Chamber 8: A cavity with a volume of approximately 100 mL, made of 316 stainless steel, with an external insulation layer and a heating jacket. The insulation layer is a heating cable 7, and the heating jacket uses an air heater 9, which can preheat and maintain the internal gas at 50-200℃. A thermocouple T6 is installed in the preheating mixing chamber for temperature measurement, and a thermocouple T5 is installed in the heating cable 7 for temperature measurement. The outlet airflow from the release module and the steam / carrier airflow from the steam generator converge here through a three-way connector and are thoroughly mixed in the cavity.

[0027] Referring to Figure 3, the migration path simulation module's function is to simulate the typical geometric path of a nuclide flowing from the leak point to the containment or environment, and to study its deposition behavior. Specific components include: Linear pipe assembly 10: composed of nine sections of borosilicate glass, each with an inner diameter of 60 mm and a length of 200 mm, connected by flanges, with a total length of 1.8 m. Each pipe section can be independently disassembled to facilitate the study of the axial distribution of deposition. The linear pipe assembly 10 is equipped with thermocouple T8 for temperature measurement. Box assembly 11: a 9.6 L volume (internal dimensions approximately 126 mm × 128 mm × 596 mm) SUS304 stainless steel box used to simulate a large process room or equipment room. Box assembly 11 is equipped with thermocouple T9 for temperature measurement. Connection and layout: The outlet of the linear pipe assembly 10 is connected to the inlet of the box assembly 11 via a reducing joint, forming a series structure of "pipe → box". The box outlet connects to the next module. Constant Temperature Environment: The entire module (pipes and enclosure) is placed inside a large, precision constant temperature chamber. This chamber is also equipped with a heating tape 7 and an air heater 9. The temperature control range of this chamber is from room temperature to 200℃, with a uniformity of ±2℃. Forced convection fans are installed on the inner walls of the chamber to ensure a uniform temperature field. Thermocouples T7 are used for temperature measurement within the chamber.

[0028] Referring to Figure 4, the nuclide collection and measurement module captures and collects all nuclides passing through the migration path simulation module in a graded and quantitative manner. Specific components (connected sequentially in the airflow direction): High-efficiency particulate filter 12: Employs a glass fiber membrane with a rejection efficiency of >99.95% for 0.3μm particles, used to collect aerosol-form nuclides. Condenser 13: Its condenser tube is made of borosilicate glass, with an effective condensation area ≥0.1m², and is circulated with constant-temperature cooling water at 4℃ to condense water vapor and nitric acid vapor, and dissolve condensable gaseous nuclides within. Collection bottle 16: Used to collect the vapor condensed by condenser 13; Primary gas scrubbing bottle 14: Contains 300mL of 0.1mol / L sodium hydroxide (NaOH) solution for chemically absorbing acidic gaseous nuclides. Secondary gas scrubbing bottle 15: Contains 300 mL of a 1:1 volume ratio hydrochloric acid-ethanol (HCl-EtOH) mixed solution, serving as a backup unit to reduce and absorb ruthenium compounds that may penetrate the primary gas scrubbing bottle 14.

[0029] Secondly, this invention discloses a method for measuring the radionuclide leakage path factor under high-level radioactive waste liquid accident conditions. Using the aforementioned measuring device, the method includes the following steps: system preparation and calibration, including: preparing a simulated high-level radioactive waste liquid 4 containing the target radionuclide; adding the simulated high-level radioactive waste liquid 4 to the sample container 3; without connecting the migration path simulation module, directly connecting the outlet of the release module to the inlet of the radionuclide collection and measurement module via a short path; heating the simulated high-level radioactive waste liquid 4 in the sample container 3 according to a predetermined temperature-controlled heating program, causing the target radionuclide to be released and absorbed by the radionuclide. The entire nuclide collection and measurement module was recovered. Quantitative analysis of the solutions and filters in each unit of the module yielded the total release amount of the target nuclide under the heating program and its release rate over time. The leakage path factor measurement experiment involved setting parameters such as the waste liquid heating program, gas phase composition ratio, carrier gas and vapor flow rate, and migration path temperature according to the target accident conditions, based on the connected release module, gas phase conditioning and mixing module, migration path simulation module, and nuclide collection and measurement module. The constant temperature chamber and gas phase conditioning and mixing module were preheated to reach and stabilize at the set temperature before starting the process. The simulated high-level radioactive waste liquid is heated by a process involving mass flow controllers and injection pumps. The release products containing the target nuclide are mixed with water vapor and / or nitric acid-water vapor mixtures generated by the steam generator in the preheating mixing chamber 8. The mixture is then sequentially passed through the migration path simulation module and the nuclide collection and measurement module until the simulated high-level radioactive waste liquid 4 is completely dry and reaches the endpoint temperature. Afterward, carrier gas and steam are introduced for a purging time. The experiment is then stopped, and various samples from the migration path simulation module and the nuclide collection and measurement module are collected. The experimental results are analyzed and calculated, including: the pipes and / or... Alternatively, the chamber may be subjected to partitioned washing and / or soaking to recover deposited nuclides. The solutions and filters in each unit of the nuclide collection and measurement module are digested. Inductively coupled plasma mass spectrometry or other appropriate analytical methods are used to quantitatively analyze the target nuclides in all samples. The total amount of deposited nuclides D recovered from the migration path simulation module and the total amount of nuclides P recovered from the nuclide collection and measurement module are calculated. Based on this, the total recovery amount T = P + D is calculated. When the total recovery rate is within a preset allowable range, the leakage path factor of the nuclide under the accident condition is obtained according to the formula LPF = P / T.

[0030] Preferably, system preparation and calibration include: calibrating the nuclide release rate by directly connecting the release module to the nuclide collection and measurement module during the heating process and bypassing the migration path simulation module, and using the total release amount and release rate curve obtained from the calibration as the benchmark for the nuclide source term in the subsequent leakage path factor measurement experiment.

[0031] Preferably, in the leakage path factor measurement experimental step, the inner surface of the pipe and / or box in the migration path simulation module is treated sequentially as follows: first, it is rinsed multiple times with an acidic solution to recover soluble nuclide deposits; then, the pipe and / or box is immersed in an oxidation-reduction eluent containing bisulfite and alkaline components and kept for a predetermined time under heating and / or ultrasonic vibration conditions to dissolve insoluble or sparingly soluble nuclide deposits; the acidic washing solution and eluent are collected separately for quantitative analysis, so that soluble and insoluble nuclide deposits in the migration path are recovered stepwise, thereby improving the accuracy of the mass balance of deposited nuclides in the leakage path factor calculation.

[0032] In a preferred embodiment of the present invention, the following steps are included: S1: System preparation and calibration a. Sample preparation: Based on the calculation results, prepare a simulated high-level radioactive waste liquid 4 (s-HLLW) containing the target nuclide (such as ruthenium nitrate).

[0033] b. Release Rate Calibration: Connect the outlet of the release module directly to the inlet of the nuclide collection and measurement module via a short path, and execute the complete heating procedure. By analyzing the total amount of nuclides collected by the nuclide collection and measurement module, determine the total nuclide release rate (ARF) and release curve under this heating procedure.

[0034] c. Gas phase calibration: Without adding waste liquid, run the gas phase conditioning and mixing module and confirm by condensation absorption method that the generated HNO3 and H2O vapor flow rates are stable and the molar ratio (e.g. NO3 / Ru, H2O / Ru) meets the preset accident scenario.

[0035] S2: Leakage Path Factor Measurement Experiment a. Parameter Setting: Based on calibration results and accident scenarios, set the following parameters: waste liquid heating program, constant temperature chamber temperature (e.g., 150℃), HNO3 and H2O supply rate of the gas phase conditioning and mixing module, and the flow rate of each carrier gas (so that the apparent gas velocity in the reaction pipeline is about 0.01m / s).

[0036] b. System preheating: Start the heating system of the constant temperature chamber and the gas phase conditioning and mixing module to reach and stabilize at the set temperature at least 1 hour in advance.

[0037] c. Experimental Procedure: Start the waste liquid heating program and simultaneously turn on all mass flow controllers and syringe pumps. Continue the experiment until the waste liquid is completely dry and reaches the final temperature (e.g., 400°C), then continue to purge with carrier gas and vapor for 30 minutes to purge any residual nuclides from the system.

[0038] d. Segmented collection: After the experiment: i. Record the liquid volume in condenser 13 and the two wash bottles respectively.

[0039] ii. Carefully disassemble the migration path simulation module. Rinse the pipes and the inner wall of the chamber three times with 50 mL of 1.0 mol / L nitric acid solution, and collect the washing solution (to dissolve soluble ruthenium species).

[0040] iii. Subsequently, the tubing and tank were immersed in fresh eluent (5 g / L K₂S₂O₅ + 0.2 mol / L NaOH) heated to 60°C for 48 hours, supplemented with ultrasonic vibration to dissolve insoluble deposits. All eluent was collected.

[0041] iv. Immerse the entire high-efficiency particulate filter 12 in nitric acid for digestion.

[0042] S3: Analytical Calculation a. Quantitative Analysis: Quantitative analysis of the target nuclides in all collected solution samples (including washing solution, eluent, condensate, absorption solution, and filter digestion solution) is performed using inductively coupled plasma mass spectrometry (ICP-MS) or other analytical methods.

[0043] b. Mass balance: Calculate the total amount of recovered nuclides. The total recovery rate (total recovered amount / calibrated total release amount) should be between 95% and 105%; otherwise, the experiment is invalid.

[0044] c. LPF calculation: Deposition amount (D) = Total amount of nuclides recovered from the migration path simulation module (pipeline + box).

[0045] Throughput (P) = Total amount of nuclides recovered from the nuclide collection and measurement module (high efficiency particulate filter 12 + condenser 13 + primary wash bottle 14 + secondary wash bottle 15).

[0046] Leakage path factor LPF = P / (P+D).

[0047] It will be understood by those skilled in the art that, unless otherwise defined, all terms used herein (including technical and scientific terms) have the same meaning as commonly understood by one of ordinary skill in the art to which this invention pertains. It should also be understood that terms such as those defined in general dictionaries should be understood to have the meaning consistent with their meaning in the context of the prior art, and should not be interpreted in an idealized or overly formal sense unless specifically defined.

[0048] For the sake of simplicity, the method embodiments are described as a series of actions. However, those skilled in the art should understand that the embodiments of the present invention are not limited to the described order of actions, because according to the embodiments of the present invention, some steps can be performed in other orders or simultaneously. Furthermore, those skilled in the art should also understand that the embodiments described in the specification are preferred embodiments, and the actions involved are not necessarily essential to the embodiments of the present invention.

[0049] As can be seen from the above description of the embodiments, those skilled in the art can clearly understand that this application can be implemented by means of software plus necessary general-purpose hardware platforms. Based on this understanding, the technical solution of this application, in essence, or the part that contributes to the prior art, can be embodied in the form of a software product. This computer software product can be stored in a storage medium, such as ROM / RAM, magnetic disk, optical disk, etc., and includes several instructions to cause a computer device (which may be a personal computer, server, or network device, etc.) to execute the methods described in various embodiments or some parts of the embodiments of this application.

[0050] Finally, it should be noted that the above embodiments are only used to illustrate the technical solutions of the present invention, and not to limit them; although the present invention has been described in detail with reference to the foregoing embodiments, those skilled in the art should understand that modifications can still be made to the technical solutions described in the foregoing embodiments, or equivalent substitutions can be made to some or all of the technical features; and these modifications or substitutions do not cause the essence of the corresponding technical solutions to deviate from the scope of the technical solutions of the embodiments of the present invention.

Claims

1. A device for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions, characterized in that, include: The system comprises, sequentially connected along the airflow direction, a release module, a gas phase conditioning and mixing module, a migration path simulation module, and a nuclide collection and measurement module. The release module includes: a sample container for holding simulated high-level radioactive waste liquid; a programmable temperature controlled furnace that covers and heats the sample container to simulate the heating process under accident conditions; and a carrier gas supply unit for conveying the released products from the sample container. The carrier gas supply unit includes a first mass flow controller connected to a carrier gas source and used to adjust the carrier gas flow rate, with its outlet connected to the upper part of the sample container to allow the carrier gas to carry the released products from top to bottom. The gas phase conditioning and mixing module includes: a steam generator for generating water vapor and / or nitric acid-water vapor mixtures; and a preheating mixing chamber connected to the outlet of the steam generator and the outlet of the release module. The preheating mixing chamber is used to... The release product gas stream from the release module is mixed and homogenized with the steam stream output from the steam generator at a set temperature to form a gaseous environment with a predetermined chemical composition and physical state. The migration path simulation module includes a pipe and / or a housing equipped with a temperature control unit to simulate the typical geometric migration path of nuclides from the nuclear facility waste liquid release point to the plant or environment under controllable temperature conditions, and to carry the flow and deposition process of the mixed gas stream. The nuclide collection and measurement module includes, in sequence along the gas stream direction, a high-efficiency particulate filter, a condenser, and at least one gas scrubbing bottle containing absorbent liquid, to sequentially intercept aerosol nuclides, condense and dissolve condensable gaseous nuclides, and chemically absorb residual gaseous nuclides, and to measure the total amount of nuclides passing through the migration path simulation module by quantitative analysis of the amount of nuclides recovered in each module.

2. The device for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions according to claim 1, characterized in that, The migration path simulation module is composed of a linear reaction pipeline assembly and a box assembly connected in series. The outlet of the linear reaction pipeline assembly is sealed to the inlet of the box assembly. The linear reaction pipeline assembly is used to simulate the migration behavior of waste liquid along linear paths such as ventilation ducts and process pipelines after release. The box assembly is used to simulate mixing, condensation and wall effects in the process room and equipment compartment.

3. The device for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions according to claim 2, characterized in that, The linear reaction pipeline assembly is composed of multiple sections of glass or stainless steel tubes with the same inner diameter, which are detachably connected by flanges. Each section can be disassembled individually to facilitate the segmented recovery of deposited nuclides. The linear reaction pipeline assembly and the box assembly are housed in a constant temperature chamber. The constant temperature chamber has the function of adjusting and maintaining a uniform temperature within a predetermined temperature range to form a stable and uniform temperature field throughout the migration path.

4. The device for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions according to claim 1, characterized in that, The steam generator includes two injection pumps for conveying deionized water and a nitric acid solution of a predetermined concentration, an atomizer for atomizing the liquid, and a steaming pipe connected to the atomizer and heated by a heating unit. The deionized water and nitric acid solution conveyed by the injection pumps are atomized into droplets in the atomizer and then enter the steaming pipe. In the steaming pipe, they are heated and vaporized to form water vapor or a nitric acid-water vapor mixture, and are output to the preheating mixing chamber under the control of a carrier gas controlled by a second mass flow controller connected to the carrier gas source.

5. The device for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions according to claim 1, characterized in that, At least one gas scrubbing bottle in the nuclide collection and measurement module contains an aqueous sodium hydroxide solution, preferably with a concentration of 0.05~0.5 mol / L, for absorbing acidic gaseous nuclides; and the nuclide collection and measurement module further includes a second gas scrubbing bottle sequentially connected to the sodium hydroxide aqueous solution scrubbing bottle, the second gas scrubbing bottle containing a hydrochloric acid-ethanol mixed solution with a volume ratio of 1:1, serving as a safeguard unit to reduce and absorb ruthenium compounds and / or other high-valence gaseous nuclides that penetrate the preceding scrubbing bottle.

6. The device for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions according to claim 1, characterized in that, The sample container is a cylindrical container made of quartz glass or Hastelloy C276, which has the characteristics of high temperature resistance and strong acid corrosion resistance. The programmable temperature-controlled heating furnace is a tubular resistance furnace with a temperature control system that can set multiple heating programs and has a temperature control accuracy of not less than ±1℃. Furthermore, the release module is equipped with a thermocouple inserted below the sample liquid surface to monitor the temperature of the simulated high-level radioactive waste liquid in real time, thereby realizing the controllable simulation of the radionuclide release rate and release stage.

7. A method for measuring the radionuclide leakage path factor under high-level radioactive waste accident conditions, characterized in that, The measuring device according to any one of claims 1-6 includes the following steps: system preparation and calibration, including: preparing a simulated high-level radioactive waste liquid containing the target nuclide, adding the simulated high-level radioactive waste liquid to the sample container; without connecting the migration path simulation module, directly connecting the outlet of the release module to the inlet of the nuclide collection and measurement module via a short path, heating the simulated high-level radioactive waste liquid in the sample container according to a predetermined temperature control heating program, so that the target nuclide is released and completely recovered by the nuclide collection and measurement module, and the nuclide is collected... Quantitative analysis of the solutions and filters in each unit of the measurement module was conducted to obtain the total release amount of the target nuclide and its release rate over time under the heating program. The leakage path factor measurement experiment included setting parameters such as the waste liquid heating program, gas phase composition ratio, carrier gas and vapor flow rate, and migration path temperature according to the target accident conditions, with the complete release module, gas phase conditioning and mixing module, migration path simulation module, and nuclide collection and measurement module connected. The constant temperature chamber and gas phase conditioning and mixing module were preheated to reach and stabilize at the set temperature before starting the waste liquid heating program and each mass flow rate. The controller and injection pump mix the release product containing the target nuclide with water vapor and / or nitric acid-water vapor mixture generated by the steam generator in the preheating mixing chamber, and then sequentially pass it through the migration path simulation module and the nuclide collection and measurement module until the simulated high-level radioactive waste liquid is completely dry and reaches the endpoint temperature. Afterward, carrier gas and steam are continued to purge for a period of time, then the experiment is stopped and various samples are collected from the migration path simulation module and the nuclide collection and measurement module. The experimental results are analyzed and calculated, including: zonal cleaning of the pipes and / or tanks in the migration path simulation module. The deposited nuclides are washed and / or soaked to recover them. The solutions and filters in each unit of the nuclide collection and measurement module are digested. The target nuclides in all samples are quantitatively analyzed by inductively coupled plasma mass spectrometry or other appropriate analytical methods. The total amount of deposited nuclides D recovered from the migration path simulation module and the total amount of nuclides P recovered from the nuclide collection and measurement module are calculated. The total recovery amount T = P + D is calculated accordingly. When the total recovery rate is within a preset allowable range, the leakage path factor of the nuclide under the accident condition is obtained by formula LPF = P / T.

8. The method according to claim 7, characterized in that, System preparation and calibration include: calibrating the nuclide release rate by directly connecting the release module to the nuclide collection and measurement module during the heating process and bypassing the migration path simulation module, and using the total release amount and release rate curve obtained from the calibration as the benchmark for the nuclide source term in the subsequent leakage path factor measurement experiment.

9. The method according to claim 7, characterized in that, In the leakage path factor measurement experiment, the inner surfaces of the pipes and / or boxes in the migration path simulation module are treated sequentially as follows: First, the pipes and / or boxes are rinsed multiple times with an acidic solution to recover soluble nuclide deposits. Then, the pipes and / or boxes are immersed in an oxidation-reduction eluent containing bisulfite and alkaline components and kept for a predetermined time under heating and / or ultrasonic vibration conditions to dissolve insoluble or sparingly soluble nuclide deposits. The acidic washing solution and eluent are collected separately for quantitative analysis, so that soluble and insoluble nuclide deposits in the migration path are recovered stepwise, thereby improving the accuracy of the mass balance of deposited nuclides in the leakage path factor calculation.

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