Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-28
Abstract： The search and localization of unknown radioactive sources is an important research topic in the field of nuclear security inspection and nuclear emergency response. In order to improve the source finding efficiency and adapt to the multi-source environment detection, a particle filtering source finding method integrating the angle of arrival is proposed. Firstly, a hardware platform combining autonomous localization and angle-of-arrival sensing is constructed to introduce position and angle information to the detector; secondly, the angle-of-arrival information is taken into account on the basis of particle filtering, which can dynamically shrink the source searching area and improve the searching efficiency; lastly, the angle-of-arrival-guided robot attitude adjustment is adopted in the path planning of the autonomous source searching, which can enhance the flexibility of the robot in searching for the source. Simulation experiments prove that this method can work correctly and effectively, and tests using radioactive sources further verify the practicality of this method for multi-source search.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-26
Abstract： [Background]: CF series fuel assemblies are the key reactor-core components of the advanced third-generation nuclear power, which are independently developed by China National Nuclear Corporation(CNNC). [Purpose]: The purpose is to analyze the friction force and rod drop performance of CF2 fuel assembly combined control rod drive line moving parts in water and air under different eccentricity. [Methods]: a 1:1 simulated fuel assembly was used in the test with an independently-developed rotatable top cap. The integration of multiple eccentric was initially implemented for scientific and accurate regulation. [Results]: The friction force and rod drop performance data in water and air at different heights and under different eccentric conditions were obtained. The total rod drop time and the time for rod reaching the buffer increased with the increase of eccentricity while the buffer time was basically constant. The fuel assembly and control rod functioned properly under the maximum eccentricity. The friction did not exceed the allowable limit. And no jamming of control rod occurred under large eccentric condition. [Conclusions]: The experimental results provide an important experimental basis for the design optimization , safe evaluation and software development of CF fuel assembly. The method can be extended to the subsequent CF3 and other fuel assembly scientific research projects.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-25
Abstract： Due to only extracting a single feature of the vibration signal of the nuclear power unit, the detection effect of the vibration fault detection method for nuclear power units is poor. Therefore, a vibration fault detection method for nuclear power units based on DBN and multi-sensor data decomposition was designed. Obtain vibration signal data of nuclear power units, smooth and fuse the obtained multi-sensor data, extract multiple features of the vibration signal of nuclear power units under the action of DBN, calculate the sensitivity index and fuzzy entropy of different features, analyze the characteristics of the vibration signal, construct a corresponding vibration fault detection model, and solve the vibration fault signal of nuclear power units. The experimental results show that in the practical application of this method, the AUC-ROC curve area is closer to 1, and the detection effect is better.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-24
Abstract： [Objective] In order to solve the problem that the existing R&D level of mobile reactors cannot meet the relevant requirements of the GB11806-2019 Regulations for the Safe Transport of Radioactive Materials, and cannot obtain transportation permits under the domestic regulatory system, [Method] In this study, according to the relevant requirements of 10 CFR 71 and §12 of the United States, the types of accidents that may be encountered in the transportation of mobile reactors were sorted out, and the most serious accident impact with the tanker truck was used as the design benchmark accident. The accident risk of a mobile reactor with an assumed power of 20 MWt under the design baseline accident is calculated. [Result] It was calculated that the probability of the design benchmark accident of the mobile reactor was 9.7×10-6/year under the condition of annual transportation, and the irradiation dose to the staff under the design benchmark accident after the reactor was cooled for one year was 810mSv. After 5 years of cooling, the radiation dose to workers under the design baseline accident was 590 mSv. [Conclusion] The accident consequences of mobile reactors under the design benchmark accident far exceed the irradiation dose limits for workers in SSG-26 Advisory Material for the IAEA Regulations For the Safe Transport of Radioactive Material and the GB18871 Safety Standards for Protection against Ionizing Radiation and Radiation Sources. Moreover, the probability of occurrence of design benchmark accidents does not meet the screening requirements of 10-7 years of over-design benchmark accidents in China. Therefore, it is necessary to take management measures, including armed escort and route planning, during the transportation of mobile reactors, to reduce the probability of design baseline accidents, and take corresponding protective measures to mitigate the consequences of design baseline accidents to meet the regulatory requirements of the nuclear safety department.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-23
Abstract： In-situ leaching is the main process of natural uranium production in China. The environmental protection problems in the process of construction, operation and decommissioning have been widely concerned. The production of wastes was introduced in this paper, and ten issues of concern were raised, included bleed ratio, drilling mud disposal, monitoring well setting, groundwater baseline determination, abnormal data determination of monitoring well, wastewater treatment, evaporation pool operation, target value determination of groundwater restoration, groundwater remediation and waste minimizaton. the present situation of technology and management was discussed, and the main task and research direction in the field of environmental protection of in-situ leaching of uranium mining were pointed out, in order to provide some suggestions for the sustainable development of in-situ leaching of uranium mining。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-23
Abstract： With the development of digital control technology, the traditional instrument and control system based on analog quantity in nuclear power plant is gradually replaced by full digital technology, and it is possible to use more complex and efficient advanced control technology. Making full use of the advantages of system information in the process of digitization of the whole plant to improve the automation level of nuclear power plant has gradually become the focus of research on pressurized water reactor control system. The control systems of Pressurizer Water Reactor (PWR) nuclear power plant are based on traditional Proportional Integral Derivative (PID) controller. Although there are studies on improving the control performance of PWR NPP control systems by advanced control algorithms, such as neural network control, fuzzy control and model predictive control, most of them only focus on the control system itself without considering the interconnection and coupling among multiple control systems. The operation task of PWR nuclear power plant needs to be coordinated by multiple control systems at the same time, and the effect of improving the overall performance by simply improving the performances of the controllers are limited. [Purpose]: To comprehensively consider the coupling effect amonge control systems, coordinate multiple control systems from the top level to optimize the overall control performances and achieve better task execution results, a setpoint decision optimization system is proposed. [Methods]: The intelligent decision system for PWR control system was optimized based on particle swarm optimization (PSO) method. The decision objective function and operation constraint conditions of the intelligent decision system were proposed. Considering the actual operation of PWR, the system optimized the setpoint offline and the intelligent decision operation was performed online according to the operation condition to provide the directions and amplitudes of the control targets for the underlying control systems. The typical operation process of the PWR NPP was taken as an example to carry out the simulation of the deigned intelligent decision-making system, and the simulation results were analyzed. [Results]: Compared with the control scheme using traditional setpoints, the ITSE (Integral of Time multiplied by the Square Error) of average coolant temperature, pressurizer level, pressurizer pressure and steam generator level decreased by 58.9%, 67.7%, 99.9% and 83.3%, respectively. The peak value decreased by 62.4%, 3.0%, 100% and 66.3% respectively. The simulation results show that the system proposed in this paper can effectively reduce the ITSE and peak value of the system. The overall control performances and safety margin of the control systems of PWR NPP are improved. In practical engineering practice, it can be combined with digital twin technology to use the characteristics of the twin that can synchronously reflect the real state of the system for more accurate online setpoint optimization, so as to achieve better control performance.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-21
Abstract： [Background]:Neural networks, with their powerful fitting capabilities, can learn the relationships between input and output variables based on large amounts of data, often serving as proxy models for physical programs in the field of engineering calculations, including nuclear engineering calculations. Neutron transport calculations, as one of the core links in neutronics simulations, often suffer from lengthy computational times. However, this issue can also be addressed by utilizing neural network models. Nevertheless, neural network models have a series of hyperparameters that need to be set, but manually adjusting these hyperparameters is laborious, repetitive, and reliant only on experience. Moreover, these hyperparameters are not reusable when solving different problems. [Purpose]: By seeking a surrogate model for VITAS, the research can provide some reference for the application of artificial intelligence in core physics calculation theory.[Methods]:This paper proposes the use of the bayesian optimization algorithm to adjust neural network hyperparameters, combined with learning rate decay and loss function optimization methods. [Results]: By fitting the key core parameters obtained from VITAS's calculation of the TAKEDA benchmark problem, the results show that the average error of the effective multiplication factor is within 150pcm, and the average error rate of the regional integral flux on the TAKEDA1 dataset is 1.72%, with a maximum error rate of 7.56%. [Conclusions]: This approach can automatically search for the optimal combination of hyperparameters for different datasets to achieve the best performance, demonstrating high flexibility, efficiency, and strong generalization.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-12
Abstract： [Background]: The primary heat exchanger (PHX) used in the 10MWt Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL), is a U-tube heat exchanger, where the shell side (hot side) contains the fuel salt from the primary loop, while the tube side (cold side) carries the coolant salt from the secondary loop. [Purpose]: This study aims to deepen the understanding and mastery of the operational characteristics of molten salt heat exchangers, and to accumulate experience in their design and operation within molten salt reactors. [Methods]: the MSRE-PHX is modeled based on the design parameters, theoretical calculations for shell and tube hear exchanger (Kern method and Bell-Delaware method), software simulation (HTRI Xchanger Suite) and computational fluid dynamics (CFD) simulation are performed, critical performance metrics, such as the heat transfer coefficient, the pressure drop, and the heat transfer power, are obtained and compared to the MSRE operation data. [Results]: The findings indicate that the discrepancies from theoretical calculations, HTRI software, and CFD simulations, are all within acceptable margins to the experimental data. Notably, the greatest variance was found with the Kern method, which showed a deviation in heat transfer quantity of about 15%, while the smallest discrepancy was observed in the overall heat transfer coefficient calculated using HTRI software, differing by merely 0.16% from the experimental data; [Conclusions]: All of the methods are suitable and applicable for designing and studying a molten salt shell and tube heat exchanger; moreover, the CFD simulation can provide fine localized details of the heat transfer and flow of the molten salt fluid. This offers substantial theoretical support and practical guidance for the future design and improvement of molten salt heat exchangers.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-12
Abstract： Accurate measurement of the control rod worth in physical tests is of great significance for the safe operation of the reactor. The research on the calculated method of multi-group space-time neutron kinetics based on hexagonal geometry was carried out, and the advanced dynamic rod worth measurement suitable for hexagonal geometry was studied. This method owns high accurate and fast, which has a wide range of applications. The results show that the deviation between the control rod worth measured by the method and the theoretical value is within 10%, which demonstrated that the calculation accuracy of this method is high. The results also show that the calculation time of the method is short, and the calculation efficiency is high.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-11
Abstract： The Dynamics beamline (D-Line), which combines synchrotron radiation infrared spectroscopy (SR-IR) and energy-dispersive X-ray absorption spectroscopy (ED-XAS), is the first beamline in the world to realize concurrent ED-XAS and SR-IR measurements at the same sample position on a millisecond time-resolved scale. This combined technique is effective for investigating rapid structural changes in atoms, electrons, and molecules in complicated disorder systems, such as those used in physics, chemistry, materials science, and extreme conditions. Moreover, ED-XAS and SR-IR can be used independently in the two branches of the D-Line. The ED-XAS branch is the first ED-XAS beamline in China, which uses a tapered undulator light source and can achieve approximately 2.5 × 1012 photons/s•300 eV BW@7.2 keV at the sample position. An exchangeable polychromator operating in the Bragg-reflection or Laue-transmission configuration is used in different energy ranges to satisfy the requirements for beam size and energy resolution. The focused beam size is approximately 3.5 μm (H) × 21.5 μm (V), and the X-ray energy range is 5–25 keV. Using one- and two-dimensional position-sensitive detectors with frame rates of up to 400 kHz enables time resolutions of tens of microseconds to be realized. Several distinctive techniques, such as the concurrent measurement of in-situ ED-XAS and infrared spectroscopy, time-resolved ED-XAS, high-pressure ED-XAS, XMCD, and pump–probe ED-XAS, can be applied to achieve different scientific goals.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-10
Abstract： [Background]: Electron cyclotron resonance heating (ECRH) is an important heating and plasma current control method for the HL-3 tokamak. Microwave inject into plasma through the launcher, which is an important part of the ECRH system. [Purpose]: Design and test the ECRH launcher system of the HL-3 tokamak. [Methods]: Design the transmission path and structure of the launcher. Simulate and calculate the effect of microwave injection. Test the transmission angle and rotation speed of the launcher, and calibrate the rotation angle of the launcher. [Results]: The optical path design of the No. 2 upper launcher has been completed. The full range response time of the equatorial launcher is less than 90ms; The full range response time of the No. 1 upper launcher is less than 190ms. [Conclusions]: The optical path design of the No. 2 upper launcher meets the requirements. The control of the equatorial launcher and the No. 1 upper launcher is precise and fast, meeting the requirements for experimental use of the tokamak.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-10
Abstract： Under high load conditions, the speed regulation system of the feedwater pump and the main/bypass valve regulation system of nuclear power units are coupled, resulting in control system oscillation, which poses a challenge to the reliability and durability of the actuator, and also affects the safe and stable operation of the nuclear power unit. Therefore, it is crucial to solve the coupling oscillation problem of control systems under high load conditions. Through in-depth analysis of the multi base fluctuation curve, it was found that the slow response characteristics of the executing mechanism and the inability to follow the response speed of the control system in a timely manner are the main reasons for the adjustment fluctuation. By evaluating the response characteristic curve of the actuator, optimizing the parameters of the valve actuator, implementing matching settings for proportional gain, speed gain, and minimum loop gain, optimizing the response characteristics in three aspects: good small signal follow-up, low large signal overshoot, and fast time response. And it improve the response and control characteristics of the actuator when it has both small and large signals. By adjusting the characteristics of the actuator, the fluctuation amplitude of the steam generator water level under high load was significantly reduced.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-09
Abstract： [Background]: In neutral beam injectors, the calorimetric target is one of the most important water-cooled components, responsible for receiving and measuring beam power. In addition, by using a built-in thermocouple array, the temperature rise at different positions of the target plate can be monitored in real-time, thereby obtaining the power density distribution of the extracted ion beam or neutral beam. [Purpose]: Develop a calorimetric target for the neutral beam injector in the HL-3 device, which can meet the requirements of target plate lifting and thermal load absorption. [Methods]: The design of the calorimetry target adopts a linear push rod mechanism to achieve lifting and lowering, and adopts a "W" - shaped target plate structure to achieve absorption of neutral beam energy. In addition, the fluid calculation module of Ansys Workbench was used to simulate the temperature distribution of the calorimetry target under full power operation. [Results]: Under full power operating conditions, the deflecting magnet is opened, and the maximum temperature rise of the calorimetry target is 526.4 degrees Celsius, which can be lowered to room temperature within half a minute, meeting the requirements for the use of the beam line. [Conclusions]: Successfully developed a calorimetric target that meets usage requirements.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-08
Abstract： The high level radioactive liquid waste (HLLW) storage building is an essential facility in the spent fuel reprocessing plant. and its stable operation is related to the production safety of the entire plant area. This article elaborates on the process system design of HLLW storage building based on design standards and engineering practice. including the HLLW storage and transportation system. circulating cooling water system. HLLW mixing system. dilution air and exhaust treatment system of the liquid waste storage tank. and so on. The overall layout form. process equipment and pipeline layout characteristics of HLLW storage building are introduced with the example of the actual project. Reference can be provided for designing of HLLW storage building and other nuclear chemical projects in the future.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-07
Abstract：蒸汽发生器传热管是压水堆核电站一回路的关键压力边界，正常运行及事故条件下的系统运行参数，直接影响着传热管结构的完整性。为进一步对传热管的结构响应特性进行研究，建立了两种主流排列方式的传热管局部模型，并分别进行了正常运行和典型事故参数下的流固耦合计算和对比分析，获得了一、二次侧流体共同作用下的传热管关键部位的载荷变化规律。结果表明，在相同的流动条件下，传热管叉排模型的应力和变形均大于顺排模型；传热管根部横截面的等效应力沿壁厚方向先减小再增大，中部横截面的等效应力沿壁厚方向呈现近似线性降低；在传热管根部区域，内、外壁面的等效应力沿轴向高度先迅速降低而后快速升高，并形成应力平台。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-06
Abstract：安全壳作为核电站的最后一道屏障，在极端情况下能够承受事故产生的内部压力，防止放射性物质泄漏。非能动安全壳冷却系统能够将安全壳内部的热量及时导出至外部循环水箱，降低事故发生时安全壳内部的温度和压力，对于保证安全壳的完整性具有重大作用。为有效评估非能动安全壳闭式冷却系统的安全性和可靠性，减少核事故的发生概率，利用数值模拟的方法对 HPR1000 中非能动安全壳闭式冷却系统进行模化分析，并对比已有实验结果进行模型验证。获得了安全壳内换热器管外传热系数、PCS 水箱温度、换热器相对高度等参数对非能动闭式冷却系统性能及启动时流动传热特性的影响规律，并以此来估算安全壳的临界值用以优化非能动安全壳闭式冷却系统，增强安全壳的安全性和可靠性。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-01
Abstract： The Tianhuan One (CAT-1: China Astro-Torus 1) is the first magnetically confined plasma device design in China using a magnetically floating dipole field magnet. According to the overall objectives and parameter design requirements of the CAT-1 device, In this paper, a simplified linear current model is used to analyze and calculate the stability of the floating magnet based on methods such as vector magnetic field, mechanical balance and dynamics. The design parameters of the floating magnet, levitation coil, and TSR (Tilt-Slider-Rotation) coil of the device, such as the overall size, spatial position and layout, current, weight, characteristic stability performance and their relationship are given. The results show that for the CAT-1 device, the optimal value of the levitation coil radius is 1.7 m, and the corresponding current is 3.49 kA for the design goal with a floating magnet current of 5MA and a height of 2.0 m. In order to achieve effective resistance and control of the floating magnet offset movement, the working area near the balance point should be limited to Δ z < 100 mm、er < 50 mm、 α< π/24.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-29
Abstract： [Background]: In pursuit of promoting the diversified development of energy cooperation demands among countries participating in the Belt and Road Initiative and address the demand for secure and efficient energy supply along the Belt and Road Economic Belt, Xi'an Jiaotong University has actively innovated and proposed a small modular fluoride-salt-cooled high-temperature advanced reactor FuSTAR. [Purpose]: Although the conceptual design of FuSTAR has been completed, the reactor's ability to operate with load tracking and its safety are still need to be verified. [Methods]: The FuSTAR system was modeled and calculated by using VITARS software for detailed thermal-hydraulic and control system modeling, and its anti-interference characteristics and load operation tracking capability were analyzed in depth. [Results]: FuSTAR has demonstrated load tracking capability without relying on an external control system, mainly due to its inherent safety features, which allow the reactor to self-stabilize and regulate under load variations. With the adoption of a constant coolant outlet temperature control scheme, the load tracking capability of FuSTAR has been further enhanced. In the tests of 10% FP load step change and 5% FP/min rate linear load rise and fall, the overshoot of nuclear reactor power is strictly controlled within 5%. [Conclusions]: Because of the negative temperature reactivity feedback and the existence of control system, the small fluoride-salt-cooled high-temperature reactor has a good load tracking ability, which fully meets the requirements of safe operation of the reactor.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-27
Abstract： [background] This study focuses on the phenomenon of bubbling in plate-type fuel assemblies within nuclear reactors. The study uses Fluent software for numerical simulation research and compares the differences between fission gas bubbles and solid bubbles, which have been previously studied. [Purpose] The purpose of the study is to investigate the effects of bubbling on temperature distribution, heat flux, and heat transfer capability in plate-type fuel assemblies. [Methods]The study employs Fluent software for numerical simulation to analyze the effects of both fission gas bubbles and solid bubbles on the thermal-hydraulic characteristics of plate-type fuel assemblies. [Results] The findings reveal that gas bubbles cause a local increase in temperature, with the heat flux around the bubbles tripling, though the overall heat flux of the fuel plate remains largely unchanged. The formation of bubbles locally enhances heat transfer capability by approximately 10%, with a 4% increase in heat flux on the bubble side. Under conditions of high flow rates, the presence of bubbles leads to a significant pressure difference across the fuel plate, causing deformation of the fuel plate and potentially leading to the blockage of the flow channel. [Conclusions] These discoveries provide significant references for the design and safety assessment of nuclear fuel plates, highlighting the importance of considering the effects of gas bubbling on thermal-hydraulic characteristics in the design and operation of nuclear reactors.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-24
Abstract： [Background]: RF power source is an important part of NNBI power system. [Purpose]: It provides RF current for RF ion source to generate plasma and maintain stable discharge. [Methods]: According to the design index, the design structure of the all-solid-state power amplifier is proposed, and the design and analysis of the RF channel buffer amplifier, pre-push amplifier, power amplifier, power synthesizer, output matching and filtering are carried out. [Results]: Finally, 150kW RF plasma discharge is realized. [Conclusions]: This circuit structure is feasible on RF ion source discharging.
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