Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-22
Abstract： The Printed Circuit Heat Exchanger (PCHE), with its intricate microchannel design, offers unparalleled heat transfer capabilities and is being considered for use in high-temperature reactors. Despite its advantages, the PCHE operates under extreme conditions, such as high temperatures and pressures, which can lead to structural failures if not properly managed. The technological readiness level of PCHEs is currently low, necessitating a detailed analysis of their failure characteristics to ensure safe and reliable operation in nuclear systems. Purpose: The purpose of this study is to design a 20 kW PCHE and investigate its potential failure modes under the demanding conditions of high-temperature and high-pressure environments. This research aims to provide a detailed understanding of the factors that could lead to failure and to propose strategies for enhancing the safety and reliability of PCHEs in nuclear applications. Methods: This study involves the design of a 20 kW PCHE and the analysis of its structural integrity under simulated high-temperature and high-pressure conditions. A partitioned homogenization method is proposed to simplify the complex microchannel structure within the heat exchanger core, facilitating numerical simulations. The study employs finite element analysis with a model consisting of 61017 elements to simulate the temperature fields and stress distributions in the solid domain. The simulations consider the effects of temperature and pressure on the micro and macro scales, providing a comprehensive view of the thermal and mechanical behavior of the PCHE. The model accounts for the thermal and mechanical properties of the materials involved, ensuring an accurate representation of the PCHE's performance under operational conditions. Results: The numerical simulations reveal significant findings. The macro temperature gradient in the solid domain reaches up to 2.7 °C/mm at the hot side inlet, indicating a high risk of thermal stress-induced failure. The temperature distribution across the cold and hot streams is non-uniform, with a maximum temperature difference of 100 °C observed at the outlets. This non-uniformity suggests potential hotspots where thermal stresses could concentrate. Additionally, the creep behavior of Alloy 617, used in the construction of the PCHE, is characterized at 850 °C. The results show that the creep life of the weld samples is significantly shorter than that of the base material, highlighting the vulnerability of the welded joints to long-term thermal exposure. The study also identifies the areas with the highest stress concentrations, which are crucial for understanding the potential for fatigue and ratcheting effects. Conclusion: The study concludes that while PCHEs offer superior thermal performance, their structural integrity under high-temperature and high-pressure conditions requires careful consideration. The identified critical areas, such as the hot side inlet and the welded joints, are particularly susceptible to failure and should be targeted for design improvements. The findings from this research provide a foundation for further studies on the safety assessment of PCHEs, emphasizing the need for robust materials and structural designs to withstand the extreme conditions encountered in high-temperature reactors. The detailed analysis presented in this study contributes to the broader understanding of PCHE failure mechanisms and supports the development of more reliable and safe heat exchanger technologies for nuclear applications. The study's comprehensive approach, integrating design, simulation, and material analysis, sets a precedent for future research in this critical area of nuclear technology.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-22
Abstract： [Background] Solving benchmark problems is a significant step in the validation of numerical simulation programs. The Experimental Breeder Reactor II (EBR-II) is a famous benchmark for sodium-cooled fast reactors (SFR), with a complicated spatial distribution of nuclide density. Therefore, simplification to the spatial distribution of nuclide density was adopted in many studies on EBR-II benchmark calculation. [Purpose] This study aims to contrast the difference between the results of the fine model and the simplified model, evaluating the rationality of the simplification. [Method] In this study, both the fine model and the simplified one were built using LoongSARAX, a neutronic numerical program for fast reactors developed by Xi’an Jiao Tong University. Some approximations were adopted in the models: one-dimensional homogenization was adopted for the half-worth driver assembly to handle its complex radial geometry and the super-assembly method was used in the cross-section generation of poison elements. [Results] The results show that in the simplified model, 1) the spatial distribution of fuel nuclide density presents strong asymmetry and strong non-uniformity, 2) calculation time spent in the simplified model is one-tenth of that in the fine model, 3) the effective multiplication factor (keff) is 1383 pcm lower than in the fine model, 4) the spatial distribution of neutron flux is lower in the center and higher in the outer core, compared to that in the fine model, 5) the maximum relative deviation between neutron flux in two models is 4.25%. [Conclusion] In summary, the simplified model has a much lower calculation cost but limited numerical accuracy in keff and neutron flux, compared to the fine model.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-21
Abstract： As nuclear energy is developed on a large scale, traditional uranium resources from mines have shown a trend of shortage, and uranium extraction from seawater is likely to be one of the most viable ways to obtain large-scale uranium resources in the future. Photocatalysis, with its advantages of low pollution, low energy consumption, and high material recycling rate, has become an important direction in the research of uranium extraction from seawater in recent years, focusing on the preparation and selection of photocatalysts. Graphite phase carbon nitride photocatalyst is relatively inexpensive and stable, making it an ideal photocatalyst; titanium dioxide photocatalyst can effectively reduce hexavalent uranium in radioactive wastewater to stable tetravalent uranium, with high activity, good stability, and environmental friendliness. This project combines the photocatalytic advantages of the several materials to prepare a magnetic composite photocatalyst made from graphite phase carbon nitride and titanium dioxide, with magnetism added to better control the catalyst in the solution and for use in seawater uranium extraction experiments. The results show that the composite catalyst has an efficiency of up to 76% in photocatalytic extraction of uranium from seawater, offering a promising new pathway for the development of uranium resources.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-19
Abstract： [Background]: 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 among 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) value of average coolant temperature, pressurizer level, pressurizer pressure and steam generator level was decreased by 58.9%, 67.7%, 99.9% and 83.3%, respectively. The peak value was 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 digital 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 2024-07-19
Abstract： [Background]: Uranium-molybdenum (U-Mo) alloy, which is applied in the heat-pipe cooled reactor (referred to as heat-pipe reactor), has the advantages of high thermal conductivity, high density of uranium and excellent irradiation performance. At the same time, U-Mo alloy as a metal material has significantly thermal expansion and irradiation swelling. And the high temperature will aggravate the irradiation swelling of U-Mo alloy and reduce the performance of the material. Hence research on swelling under high temperature of U-Mo alloy is essential in the design of this fuel. [Purpose]: The aim of this study is to comprehensively evaluate the effect of fuel swelling under high temperature of U-Mo alloy on reactor core structure. [Methods]: First, based on the irradiation data of U-Mo alloy under high temperature, a new type of swelling model considering the effect of high temperature was established. Second, a three-dimensional (3D) thermal-mechanical coupling analysis model of U-Mo alloy fuel had been set up with the use of finite element analysis (FEA) software COMSOL Multiphysics (referred to as COMSOL). Third, in order to verify the validity of the 3D FEA model, a thermal-mechanical coupling analysis was carried out considering the thermal expansion effect. The comparison and analysis with other research results showed that the thermal-mechanical coupling analysis by using COMSOL was reasonable and feasible. Then, this model was used to study the fuel swelling effect of reactor core by considering the irradiation swelling of U-Mo alloy at high temperature. The stress and deformation analysis under different burnup were carried out to evaluate the effect of fuel swelling on the core structural stability. [Results]: Under steady-state operating conditions, the core fuel of 1 kWe Kilopower heat-pipe reactor has a large deformation at the end of life (EOL) due to thermal expansion and irradiation swelling, and the maximum deformation reaches 5.28 mm. The maximum stress caused by deformation is 57.4 MPa, which is concentrated on the wall where the heat pipe is connected to the core fuel. Thermal expansion is the main factor that causes stress and deformation of fuel. As the burnup continues to deepen, the irradiation swelling of U-Mo alloy at high temperature leads to greater deformation and greater stress of the fuel. The maximum deformation of the fuel is 6.63 mm when the burnup is 0.4%, which increases by 1.69 mm compared with the calculation results considering only thermal expansion. The maximum core fuel stress reaches 85.1 MPa, which is close to the yield limit of U-Mo alloy. And the stability of fuel structure may be threatened. [Conclusions]: The results indicate that the swelling effect of U-Mo alloy at high temperature leads to more severe deformation and greater stress on the fuel. The influence of thermal expansion and irradiation swelling on the structural stability of the core at high temperature and high burnup needs to be considered in reactor fuel design. In addition, it is necessary to accelerate the irradiation test of U-Mo alloy at high temperature to optimize the irradiation behavior model.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-17
Abstract： The mixing wing plays an important role in reducing the hot spot factor of the reactor core, but there is currently limited research on the influence of the fine structure of the mixing wing on the flow field. In order to gain a deeper understanding of the influence of the characteristics of the mixing wing structure on the thermal and hydraulic performance of the reactor core, a study was conducted on the correlation between the mixing wing structure and the flow field. Firstly, through geometric automation configuration and calculation technology, parameterized and automated construction of a tearing type mixing wing structure and CFD calculation are achieved. Secondly, through orthogonal design and simulation analysis of the parameters of the mixing wing structure, the influence of the mixing wing structure on thermal parameters such as flow field pressure drop and cross flow velocity was clarified. Under the geometric structure of the article, the maximum difference in outlet pressure between different mixing wings is 1.1 kPa, which is 41% of the average pressure drop in the entire computational fluid domain. The maximum difference in cross flow velocity under different mixing wings is 1.1 m/s, which is 173% of the average cross flow velocity. The angle of the mixing wing is strongly correlated with the flow field, followed by the shape and length of the mixing wing, and a better mixing wing has been designed. Provide design basis for subsequent research and engineering application of mixing wing structures.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-16
Abstract： With the increasing requirements for the accuracy of numerical calculations in reactors, 0more and more attention has been paid to multi-physics coupling calculation in nuclear reactor analysis. [Purpose]: As one of the main calculation methods in industry, the study of the neutronics and thermal-hydraulics coupling calculation method which is suitable for the deterministic two-step method has a clear value for industry. [Methods]: Based on the two-step codes DRAGON/DONJON and the subchannel code COBRA-EN, a neutronics and thermal-hydraulics coupling code system based on a unified framework has been developed, and the coupling code system has been validated around the Virtual Environment for Reactor Applications (VERA) series of benchmark problems 6 and 7, which were proposed by the Consortium for Advanced Simulation of LWRs (CASL), in the United States. [Results]: The results show that the error of keff for Problem 6 is within 100×10-5, and the radial fission rate of the component is within ±1%, the distribution trends of fuel temperature and coolant temperature agree well with reference values; While the computational error of the critical boron concentration for Problem 7 is within 20×10-6, and the root-mean-square error of the power distribution is 0.86%, the coolant temperature at the core outlet differs from the reference value within ±278.15K .[Conclusions]: The computational capability of the coupling code system is verified.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-15
Abstract：随着化石燃料逐渐枯竭，研制商业化的聚变反应堆成为解决能源问题的途径之一，而为了实现聚变堆氚自持运行，从聚变堆增殖包层中提取氚的技术有着重要研究意义。从经济性和可实现性角度出发，液态锂铅包层具有很大的优势和发展前景。本文对国际热核聚变实验堆（ITER）以及欧盟聚变示范堆（DEMO）所提出的液态锂铅包层中的氚提取技术，包括真空渗透法（PAV）、气液接触法（GLC）、真空筛板法（VST）三种氚提取工艺进行了综述，归纳总结了各个工艺的优劣势、所需辅助系统、技术问题等，并对其进行了性能比对，以期对聚变堆氚提取系统的设计研究提供思路和参考。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-15
Abstract： [Background]: Molten salt reactor is one of the six internationally recognized and recommended fourth generation reactors , which is different from conventional solid-state nuclear fuel reactors. It is necessary to analyze the relationship between 7Li abundance and nuclear critical parameters in order to do a good job in core design management and nuclear safety supervision. [Purpose]: This study modeled a molten salt reactor with reference to engineering practice, and used software simulation calculations to analyze the impact of different 7Li abundance fuel salts on the reactivity of the molten salt reactor, and to analyze the changes in nuclear critical parameters. [Methods]: By iterative calculation, quickly and accurately find the 7Li abundance value at the critical state of the molten salt reactor core. [Results]: The final conclusion is that the reactivity of the molten salt reactor increases with the increase of fuel salt 7Li abundance, and the rate of change in reactivity of the molten salt reactor is also related to 7Li abundance. [Conclusions]: Based on the analysis results of this study, conduct in-depth exploration, summarize relevant management requirements from the perspective of laws and regulations, and propose key points for attention from the perspective of review and management.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-14
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 2024-07-13
Abstract： [Background]: As the commercial application of sodium-cooled fast reactors continues to develop, there is a need to develop an integrated analysis code for sodium-cooled fast reactors in order to enhance the capability for safety assessment. [Purpose]: Sodium boiling is a critical phenomenon prior to severe accidents in sodium-cooled fast reactors, and accurate prediction of thermodynamic parameters of the coolant at the onset of sodium boiling is essential for accident analysis. [Methods]: This paper introduces the development of the integrated analysis code ISAA-Na for sodium-cooled fast reactors, focusing on the core thermal-hydraulic model. The ISAA-Na code was used to simulate and analyze the loss-of-flow experiment CABRI-BI1. [Results]: Comparison with experimental data shows that ISAA-Na provides more accurate predictions of coolant temperature and pressure prior to boiling compared to other codes such as SAS4A, ASTEC-Na, and SIMMER. However, after the onset of boiling, overestimations of bubble growth rate and two-phase interface movement were observed due to the lack of fuel melting and cladding failure models. [Conclusions]: Overall, the application of the multi-bubble liquid slug boiling model in ISAA-Na for the analysis of the CABRI-BI1 experiment is deemed reasonable and reliable.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-12
Abstract： [Background] The imperative need for high-performance propulsion systems in deep space exploration missions has led to a focus on improving the design of nuclear thermal propulsion reactors (NTPRs). The existing methods for designing NTPRs have been identified as lacking in systematic and integral approaches. [Purpose] The purpose of this study is to propose a novel multi-objective optimization design method for NTPRs to achieve a core design that offers high thrust, high specific impulse, extended service life, and reduced weight. [Methods] The methodology involves several key steps: Constructing a heat transfer model between assemblies based on their thermal interaction characteristics. Integrating this model with the flight performance model of the nuclear rocket and the two-dimensional criticality model of the assemblies. Proposing a multi-objective parameter screening method that combines the aforementioned models for coupled iterative calculations to optimize the core layout. Ensuring that the design meets comprehensive standards in thermal engineering, flight performance, and neutron physics while minimizing the core mass. Utilizing the open-source Monte Carlo software OpenMC to perform detailed three-dimensional neutronics calculations and conduct a comprehensive assessment of the reactor's criticality, safety, and burnup performance. [Results] The study's findings demonstrate that the low-enriched uranium (LEU) NTPR conceptual design, developed using the proposed method, has preliminarily satisfied the design criteria for high thrust, high specific impulse, long service life, and lightweight. [Conclusions] The results suggest that the proposed method for NTPR core design is effective in meeting the demanding requirements of future manned deep space exploration missions, offering a promising direction for further research and development in this field.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-12
Abstract： [Background]: Radiation shielding is an important method to ensure the environmental safety of personnel and nuclear facilities in the nuclear industry. As it usually consumes a long time in single simulation calculation, the optimization design of radiation shielding is a classical expensive optimization problem. [Purpose]: In order to reduce consumption time and improve optimization efficiency in radiation shielding optimization, a dynamic radial basis surrogate model based on particle flight sample update strategy is proposed. [Methods]: A radial basis neural network was first used to build the initial surrogate model of the actual objective function. The surrogate model was then globally searched for optimality by a differential evolutionary algorithm, and new sample points were selected to join the original sample points based on the result of the surrogate model search and the particle flight sample update strategy. Finally, the surrogate model was updated based on the new set of sample points and iterated until the convergence condition was satisfied. Since this method controlled the flight speed of the original sample point to the random sample point and the optimal predicted sample point based on the fitting accuracy of the surrogate model, the adaptive balance between the global exploration and the local exploration of the dynamic surrogate model could be achieved. In order to verify the effectiveness of the method, the proposed method was applied to 12 numerical test functions and the optimization design for radiation shielding of marine reactors, and the calculation results of other optimization methods were compared. [Results]: The results show that for numerical test functions, the proposed method has significant advantages in search accuracy, search efficiency, and algorithm robustness. For the radiation shielding optimization, the neutron transmittance obtained by the proposed method is 48% and 8% of the other two methods, and the number of required sample points is 25% of the static surrogate model. [Conclusions]: The research results indicate that by using the dynamic surrogate model based on the particle flying algorithm, the number of sample points needed to solve the expensive optimization problem is greatly reduced. The dynamic radial basis surrogate model with particle flight is an effective method for radiation shielding optimization.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-12
Abstract： [Background] The Coefficient of heat transfer (HTC) is affected by many factors, and there are some problems such as unclear physical model and lack of experimental data. The traditional empirical relation is difficult to meet the high precision numerical calculation. Machine learning algorithms can effectively solve complex nonlinear problems, but some results do not conform to physical laws. [Purpose] The aim of this study is to propose a physical information machine learning algorithm model that can calculate thermal parameters more accurately. [Methods] First, HTC experimental data were collected under a circular tube. Secondly, The HTC model was developed by combining the Jens-Lottes formula and Thom formula with MLP, BPNN and RF. Finally, six algorithms in the HTC models were evaluated and compared with empirical correlations. [Results] In the HTC model, the calculation accuracy of Jens-Lottes formula combined with RF is the highest, it has good extrapolation. At the same time, using the physical information machine learning algorithm could effectively improve the computational accuracy of the physical model. [Conclusions] The physical information machine learning algorithm could provide a method to build a high precision calculation model of HTC.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-12
Abstract： Radiation shielding is an important method to ensure the environmental safety of personnel and nuclear facilities in the nuclear industry. As it usually consumes a long time in single simulation calculation, the optimization design of radiation shielding is a classical expensive optimization problem. [Purpose]: This study aims to reduce the number of sampling points required in the radiation shield optimization design and improve the efficiency of intelligent optimization algorithms. [Methods]: A radial basis neural network was first used to build the initial surrogate model of the actual objective function. The surrogate model was then globally searched for optimality by a differential evolutionary algorithm, and new sample points were selected to join the original sample points based on the result of the surrogate model search and the particle flight sample update strategy. Finally, the surrogate model was updated based on the new set of sample points and iterated until the convergence condition was satisfied. Since this method controlled the flight speed of the original sample point to the random sample point and the optimal predicted sample point based on the fitting accuracy of the surrogate model, the adaptive balance between the global exploration and the local exploration of the dynamic surrogate model could be achieved. In order to verify the effectiveness of the method, the proposed method was applied to 12 numerical test functions and the optimization design for radiation shielding of marine reactors, and the calculation results of other optimization methods were compared. [Results]: The results show that for numerical test functions, the proposed method has significant advantages in search accuracy, search efficiency, and algorithm robustness. For the radiation shielding optimization, the neutron transmittance obtained by the proposed method is 48% and 8% of the other two methods, and the number of required sample points is 25% of the static surrogate model. [Conclusions]: The research results indicate that by using the dynamic surrogate model based on the particle flying algorithm, the number of sample points needed to solve the expensive optimization problem is greatly reduced. The dynamic radial basis surrogate model with particle flight is an effective method for radiation shielding optimization.
Peer Review Status:
Awaiting Review
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-11
Abstract： [Background] As a typical representative of passive safety technology, the automatic depressurization system ( ADS ) accelerates the reactor primary loop depressurization after the accident, connects the high-pressure, medium-pressure and low-pressure safety injection systems, and maintains the core cooling. [Purpose] In order to study the opening and closing characteristics of the pressure relief valve of the automatic depressurization system and its influence on the reactor systems. [Methods] Based on the system analysis program, China Advanced Pressurized Water Reactor is taken as the research object, and the typical ADS trigger accident is taken as the initial event. The different opening speeds of the first three valves of ADS and the closing conditions of the fourth pressure relief valve of ADS are simulated, and the response of each system under different working conditions is analyzed. [Results] The results show that, the opening speed of the ADS1-3 valve cannot significantly affect the pressure relief characteristics of the primary loop; the ADS-1 pressure relief valve uses a quick opening method which helps the sprinkler reach the stable critical jet state faster; the ADS-2/3 pressure relief valve uses a slow opening method which is helpful to avoid "sharp" peak points in the pipeline flow curve and reduce the impact of the spraying process on the pipeline and sprinklers while satisfying the economic principle; ADS-4 is crucial for small break LOCA accident and also necessary for the injection of pressure tank at the later stage of the accident. [Conclusions] Through the simulation analysis of the opening and closing characteristics of the ADS pressure relief valve, it provides a reference for the design of the automatic depressurization system, and also provides theoretical and data support for the safety analysis of advanced nuclear power plants.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-09
Abstract： With the increasing requirements for the accuracy of numerical calculations in reactors, more and more attention has been paid to multi-physics coupling calculation in nuclear reactor analysis. [Purpose]: As one of the main calculation methods in industry, the study of the neutronics and thermal-hydraulics coupling calculation method which is suitable for the deterministic two-step method has a clear value for industry. [Methods]: Based on the two-step codes DRAGON/DONJON and the subchannel code COBRA-EN, a neutronics and thermal-hydraulics coupling code system based on a unified framework has been developed, and the coupling code system has been validated around the Virtual Environment for Reactor Applications (VERA) series of benchmark problems 6 and 7, which were proposed by the Consortium for Advanced Simulation of LWRs (CASL), in the United States. [Results]: The results show that the error of keff for Problem 6 is within 100 pcm, and the radial fission rate of the component is within ±1%, the distribution trends of fuel temperature and coolant temperature agree well with reference values; While the computational error of the critical boron concentration for Problem 7 is within 20 ppm, and the root-mean-square error of the power distribution is 0.86%, the coolant temperature at the core outlet differs from the reference value within ±5°C .[Conclusions]: The computational capability of the coupling code system is verified.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-09
Abstract： Currently ，Uranium molybdenum alloy (U-Mo) is a very popular fuel used in research reactors, space reactors, and small reactors for special purposes. During the irradiation process, the microstructure of the U-Mo fuel will undergo a series of changes, which may affect the fuel performance during reactor operation. These changes mainly include ：the formation of the interaction layer between the U-Mo fuel core and the matrix, the release of fission products (mainly the release of fission gas), and the refinement of the grain size for U-Mo fuel with high burnup. This article summarizes the main characteristics of the above changes and the latest research progress for them. At the same time, it also proposed the development trends in the study of microstructural changes after fuel irradiation in today's world where advanced detection technology has made significant progress.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-04
Abstract： Under severe Ocean environment, the liquid level inside the in-vessel pressurizer will fluctuate greatly, threatening the safe operation of the reactor. To study the response characteristics of the liquid level inside the in-vessel pressurizer under severe ocean conditions, the fluctuation forms of the liquid level pressure under various motion forms are analyzed by means of experiments. The results show that the swinging motion affects the pressure signal because the linear distance between the liquid surface and the measuring point changes during the swinging process. The swaying(surging) motion that influence the pressure measurement are derived from the tendency that the fluid will periodically impact the wall and liquid surface due to inertia. The influence of heaving motion on pressure measurement mainly comes from the acceleration change in the vertical direction, which periodically causes the phenomenon of weightlessness.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-02
Abstract： [Background] CF series fuel assemblies are the core components of nuclear power independently developed by China National Nuclear Corporation(CNNC), which together with the drive mechanism, constitute an important part of a large pressurized water reactor. And they are related to the stability and safety of nuclear reactor operation. [Purpose] The development of any fuel assembly and related equipment must undergo friction and rod drop test to verify its structural integrity and the property. The experimental study has been carried out for analyzing the friction and rod drop performance of the drive line’s moving parts of the self-developed CF2 fuel assembly under water and air working conditions with different eccentricity. [Methods] The traditional friction and rod drop performance test can only be carried out with small eccentricity to obtain the test data such as friction force and rod drop curve. It can't be achieved to analyze and study the performance with multiple eccentricities, especially the larger eccentricity. And there are no more studies on comparative mechanical analysis of the friction force of the driving mechanism through full displacement and in the two media of water and air. a 1:1 simulated CF2 fuel assembly was used in the test with an independently-developed rotatable top cap. The integration of multiple eccentricity was initially implemented for scientific and accurate regulation. The method to study the performance of the driving mechanism was optimized. The friction force and rod drop performance data were obtained through the experiments of the driving mechanism in water and air through full displacement and under different eccentric conditions. [Results] The control rod has been uniformly raised from the lowest to the highest position in the air at an ambient temperature of 22°C and with an eccentricity of 4.67 mm, 6.0 mm and 9.4 mm, respectively. When the eccentricity is 4.67 mm, the friction force at the lowest and highest positions is 26.5 N and 32.1 N, respectively. And when the eccentricity is 9.4 mm, the friction force at the lowest and highest position is 37.1 N and 39.2 N, respectively. Then the control rod has been uniformly raised from the lowest to the highest position in the static water at a temperature of 18°C and with an eccentricity of 4.67 mm, 6.0 mm and 9.4 mm, respectively. The friction force at the lowest position and the highest position is 54.8 N and 39.5 N respectively when the eccentricity is 4.67 mm. When the eccentricity is 9.4 mm, the two force values in the water are 62.8 N and 44.1 N, respectively. As the control rod is gradually raised, the friction force in the air increases much more than that in static water. The maximum value in the air occurs at the highest position, while the maximum value in static water occurs at the lowest position. By comparing and analyzing the rod drop performance and parameters like velocity, displacement and vibration, it can be seen that with the continuous increase of the eccentricity, the friction during rod operation also increases correspondingly, and the maximum speed during rod drop gradually decreases. The data shows that the impact of the various eccentricities on the buffer time is small and the value can be basically consistent. The minimum rod drop time occurs when the eccentricity is 0 mm. The time for control rod reaching to the buffer port is 1.049 s, and the time reaching to the bottom is 1.477 s. [Conclusions] The centering work has been successfully completed by using the self-developed top cap, which changes the traditional drive mechanism centering way. It’s an innovation for fast centering with multiple eccentricities, and it greatly improves the test efficiency and reduces the test cost. The numerical comparison of friction force between in the air and static water through full displacement and under multiple eccentricities has been carried out, which expands the research scope in this field. Through multiple friction and rod drop performance tests, the fuel assembly and control rod run well, the friction does not exceed the limit value, and no jamming of the rod occurs under the maximum eccentric condition. It can verify the rationality of the design of CF2 fuel assembly, and provide an important test basis for the design, safety evaluation and software development of this series fuel assemblies.
Key words CF2 fuel assembly, Rod Drop Performance, Friction
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