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.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-24
Abstract： Background : Subchannel analysis of fuel assemblies is critical for the development of lead-bismuth reactors. Purpose : The purpose of the study is to modify and optimize the COBRA subchannel program to make it suitable for lead-bismuth reactors and validate its performance. Methods : Modifications were made to the COBRA subchannel program, involving adjusting physical properties, convective heat transfer models, friction models, and turbulence mixing models. The performance of the modified program was then compared to experimental data and numerical calculations. The influence of mass flow rate on calculation accuracy was analyzed, and it was found that the error of the sub-channel increased after the mass flow rate decreased. To optimize results over a wide range of mass flow rate conditions, an optimization method based on a subchannel model and coupled with a neural network was proposed. Results : The results demonstrate that the modified subchannel program performs well under experimental conditions, with an error of no more than 5% compared with experimental results and no more than 3% compared with FLUENT results. The application of neural networks is found to improve accuracy and reduce errors by an order of magnitude. Conclusions : The optimized subchannel analysis method, derived from the modifications and neural network coupling, can accurately predict outlet temperatures for lead-bismuth reactors under a wide range of mass flow rate conditions. This method provides valuable guidance for the design of such reactors.
Peer Review Status:
Awaiting Review
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-22
Abstract： Abstract [Background]: The defects generated during the working process of metal materials have a significant impact on their performance. For example, the radiation-induced embrittlement and hardening of RPV steels are a factor of concern, which hinders the life extension of the RPV. Annealing treatment is applied to alleviate irradiation-induced precipitates and defects and recover RPV’s mechanical properties in the past few decades to extend the in-service lifetime of the RPV. Unfortunately, this conventional method generally requires a high treatment temperature and long operation time, inevitably wasting considerable energy due to the huge size of the RPV. Recently, as a more convenient and energy-saving method, the repair of metal defects by electropulsing treatment (EPT) has been developed.
[Purpose]: Design and construct a device for EPT processing of samples. The repairs of defects in electron irradiated and deformed iron and RPV steel after EPT has been investigated by using positron lifetime spectroscopy.
[Methods]: Electron irradiated pure iron and RPV steel samples were subjected to multi parameter EPT, and the changes in defects of the samples with EPT were characterized by positron lifetime spectroscopy. In addition, the mechanical properties of pure iron tensile samples were characterized by micro Vickers hardness, and the defect information was characterized by positron lifetime spectroscopy to investigate the relationship between macroscopic properties and microstructure.
[Results]: The defects introduced by electron irradiation in pure iron and RPV steel samples gradually recover after EPT and exhibit similar patterns to annealing treatment. After stretching, the number of defects in pure iron samples increases, leading to an increase in Vickers hardness. EPT can restore defect and reduce Vickers hardness.
[Conclusions]: The EPT equipment and method used in the experiment can indeed recover the defects of metal samples. The samples treated by EPT are characterized by positron lifetime spectroscopy, and defects generated by irradiation or deformation in pure iron and RPV steel can be partially repaired through EPT. The effect of defect repair is not only related to the initial state of the sample, but also to EPT’s parameters. As a new non-destructive testing method, positron annihilation is expected to provide a criterion for material damage or defect repair under the action of pulse current, which can conveniently, quickly, and sensitively detect the defect state of actual working components.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-15
Abstract： Background Reactor activation-burnup calculation is aan important crucial part component of reactor analysis, which is the iterative calculation by coupling the criticality program and point burnup programinvolving an iterative process that combines criticality programs with point burnup programs. Purposes LightAB (Light Activation and Burnup) is a novelnew lightweight, general-purposelight general-purpose activation-burnup program designed for this purpose. Methods LightAB used utilized burnup databases based on ORIGEN-2 and ORIGEN-S, the ORIGEN-2 and ORIGEN-S-based burnup databases, implementingand employed a Chebyshev rational approximation (CRAM) algorithm for rigid accurate burnup systems. LightAB supporteded point burnup calculation in decay mode, constant flux mode and constant power mode. MoreoverThe programs was well-structured, LightAB was highly packaged and divided intoconsisting of a solver module, an I/O module, and a burnup chain module. LightAB taked used nuclide as the basic storagefundamental unit of storage, and the input module of LightAB specifiedd physical quantities such as burnup database path and sub-burnup step division. Results The 237Np decay problem and the fixed-flux irradiation problem of Zr are calculated to verify the correctness of LightAB, where its results is consistent with that of ORIGEN2.1. LightAB and RMC programs are coupled and used to calculate the Pressurized Water Reactor (PWR) cell burnup model, the PWR assembly burnup model, and the OECD/Nuclear Energy Agency (OECD/NEA) fast reactor burnup model, and the results are consistent with that of RMCLightAB’s accuracy has been demonstrated through the successful calculation of the decay of 237Np and the irradiation of Zr under fixed-flux conditions, yielding results consistent with ORIGEN 2.1. By coupling with RMC programs, LightAB has been utilized in calculating various reactor burnup models, including Pressurized Water Reactor(PWR) cell, PWR assembly, and OECD/Nuclear Energy Agency(OECD/NEA) fast reactor models, with results being consistent with RMC calculations. Conclusions LightAB can be used in the simulation of radiation production of transplutonium isotopes.has shown promising application prospects in the irradiation production of transplutonium isotopes compared with RMC simulation calculations.
Peer Review Status:Awaiting Review
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-15
Abstract： CAT-1 (China Astro-Torus 1) is a levitated dipole field magnetic confinement device, which mainly used for dipole plasma physics experiments, requiring a central floatin
g superconducting coil to be stably levitated for at least 5hr without cooling or power supply. In this paper, a levitation control system of coupling superconducting levitation coil and floating coil is designed. In order to ensure the stable levitation of 1200 kg, 5 MA floating magnet, Simulink model of the control system is established and simulated. Based on Routh-Hurwitz stability criterion, the influence of PID（Proportion-Integral-Derivative） control strategy on stability control is studied. The selection range of stability control parameters is determined: Under ideal conditions, delay time of the PD（Proportion-Derivative） control system is 0.0463s, rise time is 0.1545s, peak time is 0.6283s, adjustment time is 0.0848s, and overshot δ=1.6. The results show that PID can restore the levitated superconducting ring to the preset balance position in a short time, and the load of the circuit can be greatly reduced by adopting the appropriate starting mode. The results provide key technical support for the design and development of levitated superconducting dipole field devices.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-14
Abstract： Background : The quantification of uncertainty has become a common requirement in reactor physics analysis, with the covariance data of nuclear data serving as the foundational data for conducting uncertainty quantification. Purpose : This study aims to develop a covariance data generation module, covar_calc, in the nuclear data processing software NECP-Atlas to produce continuous energy covariance data for Monte Carlo programs and multi-group covariance data for deterministic programs. Methods : Covar_calc module was developed to process all covariance data provided in evaluated nuclear databases, according to the different storage formats of nuclear data and different computational methods. This included the covariance data of various parameters such as average fission neutron multiplicities, cross sections, angular distributions of secondary particles, fission spectrum, resonance parameters, and neutron activation cross-sections. Results : A comparative verification is carried out with the covariance data production module in the nuclear data processing software NJOY21 to demonstrate the accuracy of the multi-group covariance databases produced by NECP-Atlas and the maximum bias is less than 0.1%. Utilizing Monte Carlo calculation code NECP-MCX and uncertainty analysis code NECP-UNICORN, sensitivity coefficients for different benchmarks are calculated. By incorporating both continuous energy covariance databases and multi-group covariance data, and utilizing the Sandwich formula, the final uncertainties are computed. The results indicate that the uncertainties calculated using the multi-group covariance data generated by NECP-Atlas exhibit comparable accuracy to those obtained with NJOY21. Conclusions : The precision in creating both continuous energy covariance and multi-group covariance meets the requirements for usage in Monte Carlo programs and deterministic programs, validating the efficacy of covar_calc module within NECP-Atlas for uncertainty quantification in reactor physics analysis.
Peer Review Status:Awaiting Review
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-09
Abstract： Background : Method of Characteristics (MOC) is widely used in high-fidelity numerical simulation because of its powerful geometric processing ability and its ability to take into account calculation cost and accuracy. In addition to MOC, the common transport calculation methods also include Collision Probability Method (CP) and Interface Current Method (IC). Purpose : In this paper, the comparative analysis of MOC, CP and IC from method theory and numerical calculation is particularly important for evaluating their ability in pin-by-pin calculation. At the same time, in MOC calculation, different parameter selection will affect the calculation cost and calculation accuracy, so it is necessary to carry out sensitivity analysis to find the best parameter. Methods : In this paper, the three calculation methods are compared from the principle, and then the numerical calculation and the preliminary analysis of the sensitivity of MOC parameters are completed based on the 2D C5G7-MOX reference problem. Results : The calculation time and memory cost of MOC are 23.9min and 37.5MB respectively, and the relative error between MOC and reference solution is only 6.04×10-4. The computing time of CP and IC is 56.7 times and 15.6 times that of MOC, and the memory cost is 407.7 times and 32.8 times that of MOC, respectively. Based on the sensitivity analysis of MOC parameters, a set of suggested parameters are given: grid division is 6×6, pole Angle is GAUS and number is 2, and azimuth Angle is 30. The calculation time of this set of parameters is 45.4min, the memory cost is 264.7MB, the relative error is 5.9×10-5, and the normalized RMS error is 0.00255. Conclusions : The results show that MOC is superior to CP and IC in accuracy, efficiency and memory cost. Through the parameter sensitivity analysis, it is found that the grid division has the greatest influence on the calculation memory cost and calculation time, and the choice of polar Angle has the greatest influence on the calculation accuracy.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-04-08
Abstract： The safety assessment of high-level radioactive waste repositories requires a high predictive accuracy for radionuclide diffusion and a comprehensive understanding of the diffusion mechanism. In this study, a through-diffusion method and six machine-learning methods were employed to investigate the diffusion of ReO4– , HCrO4– , and I– in saturated compacted bentonite under different salinities and compacted dry densities. The machine-learning models were trained using two datasets. One dataset contained six input features and 293 instances obtained from the diffusion database system of the Japan Atomic Energy Agency (JAEA-DDB) and 15 publications. The other dataset, comprising 15,000 pseudo-instances, was produced using a multi-porosity model and contained eight input features. The results indicate that the former dataset yielded a higher predictive accuracy than the latter. Light gradient-boosting exhibited a higher prediction accuracy (R2= 0.92) and lower error (MSE = 0.01) than the other machine-learning algorithms. In addition, Shapley Additive Explanations, Feature Importance, and Partial Dependence Plot analysis results indicate that the rock capacity factor and compacted dry density had the two most significant effects on predicting the effective diffusion coefficient, thereby offering valuable insights.
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